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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 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

A new tubular ironless linear induction motor using non-interrupted winding method

The ironless linear induction motor device can be used for any application which requires a large motion span or rotation restrictedly by the ironless feature due to the strong magnetic field environment. One of the possible applications is ITER diagnostics shutters. The tokamak reactor systems have a severe radiation issue during the operation. The traditional Kapton-insulated wires have significantly lower radiation resistance than an expected large radiation dose at ITER Motion Stark Effect (MSE) shutter location. Furthermore, the traditional single-sided linear induction motor winds the wire to make the N-turn pitch coils independently and then connect them in series to make a one-phase winding. Therefore, it is a big challenge to make so many radiation-protection connectors in such a limited space as inside the vacuum vessel of fusion reactors. The approved mineral-insulated kinds of cables need a large bend radius for the end turns, so the traditional single-sided linear motor with two-layer three-phase windings need to be upgraded to the tubular linear motors. The new proposed tubular ironless linear induction motor can not only overcome the large bend radius issue with the mineral-insulated cables but also allow the motor coils wound continuously around the tubular-shaped stator, forward and backward N times to make a N-turn winding for each phase, in order to resolve the radiation protection issue with the winding connection. The 3D transient Maxwell electromagnetic model with a tubular three-phase linear induction motor has been analyzed. Furthermore, this paper will present the detailed design and winding method of ironless linear induction motor actuators, and results of the 3D transient numerical simulation of the actuator performance.

Ironless↗
NASA NTRS1972

Post impact behavior of mobile reactor core containment systems

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

Puthoff, R. L.↗
DOE OSTI2015

Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU analyses. Additionally, current analyses suggest that the NUREG-1465 release fractions are conservative by about a factor of 2 in terms of release fractions and that release durations for in-vessel and late in-vessel release periods are in fact longer than the NUREG-1465 durations. It is currently planned that a subsequent report will further characterize these results using more refined statistical methods, permitting a more precise reformulation of the NUREG-1465 alternative source term for both LBU and HBU fuels, with the most important finding being that the NUREG-1465 formula appears to embody significant conservatism compared to current best-estimate analyses. ACKNOWLEDGEMENTS This work was supported by the United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The authors would like to thank Dr. Ian Gauld and Dr. Germina Ilas, of Oak Ridge National Laboratory, for their contributions to this work. In addition to development of core fission product inventory and decay heat information for use in MELCOR models, their insights related to fuel management practices and resulting effects on spatial distribution of fission products in the core was instrumental in completion of our work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Micro- and macro-structural heterogeneities in 316L stainless steel prepared by electron-beam melting

Highlights: • Electron beam melting manufactured parts has a complicated thermal history. • The microstructure varies with the position in the sample. • Pronounced molybdenum segregation at the top disappears at the bottom. • Fast molybdenum diffusion in cell boundaries as well as grain boundaries. - Abstract: This is a study of the micro- and macrostructural variations in samples of stainless steel with the overall composition of the grade 316L, produced using electron beam melting. Electron beam melting is one of the processing methods under consideration for manufacturing some of the International Thermo- Nuclear Experimental Reactor In-Vessel components. Therefore further studies of the homogeneity of the material were conducted. Electron beam melting results in a complicated thermal history of the manufactured part giving a significant impact on the microstructure. A cellular structure that is often observed in samples prepared by selective laser melting was found in the top layers of the specimens. Further down, the structure changed until the cellular structure was almost non-existing, and the grain boundaries had become more pronounced. This revelation of a heterogeneous structure throughout the entire part is crucial for large-scale industrial applications like the Thermo- Nuclear Experimental Reactor to make sure that it is understood that the properties of the material might not be the same at every point, as well as to assure that the correct post-treatment is done. It is also exposed that a significant part of this change is due to molybdenum redistribution inside the sample when it diffuses from the cell boundaries into the cells, and into bigger agglomerates in the grain boundaries. This diffusion seems not to affect the microhardness of the samples.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Preliminary Design of Ironless Linear Induction Motors for ITER MSE Shutter Actuators

In this article, the shutters need to be closed or opened during the operation and calibration period time for the fusion diagnostic systems, such as the International Thermonuclear Experimental Reactor (ITER) motional stark effect (MSE). Standard electric motors and actuators will not work in a strong magnetic field environment due to the presence of the magnetic field of fusion reactors. The innovative linear induction motor (LIM) with an ironless feature overcomes this kind of challenge and can be used for any application that requires controlled motion with a large stroke in the magnetic field environment. It consists of a high-electrical conductivity plate like copper and three-phase motor windings with nonferrous or stainless-steel stators. The conductor plate attached with the shutter is simply driven by the underneath three-phase linear windings to achieve linear motion. The motor drive can be controlled remotely by a controller using the electrical connection, so no sensitive electronic components are located in the harsh environment where the motor itself is located. The design requirements and test facility have been described. Several 3-D transient Maxwell electromagnetic (EM) models with different locations of three-phase linear motor windings and shutter stators have been analyzed and evolved to meet any applications allowed in the harsh environment inside the vacuum vessel of fusion reactors. The preliminary design results are presented in this article.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Load planning in the dark: RH-TRU waste container load planning at Sandia National Laboratories/New Mexico - 15116

Sandia National Laboratories (Sandia) performed several sets of experiments during the 1970's and 1980's with fresh and irradiated mixed oxide (MOX) and uranium oxide (UO{sub 2}) materials designed to simulate various fuel responses under varying reactor conditions. These research materials were from light water and breeder mixed oxide reactors. The experimental vessels and extraneous piece parts resulting from the experiments were identified as eligible for disposal at the Waste Isolation Pilot Plant (WIPP) as remote-handled (RH) transuranic (TRU) waste. Once Sandia received the approval to dispose at WIPP, discussions began with the Carlsbad Field Office (CBFO) and Nuclear Waste Partnerships, LLC (NWP) to provide Central Characterization Program (CCP) support for certified acceptable knowledge, visual examination, and dose-to-curie for the repackaging effort. In addition to the WIPP support, Sandia was required to prepare a load plan describing the configuration of the parent containers and how they could be repackaged to meet WIPP and other regulatory requirements. However, all of these containers were packaged as material twenty (20) or thirty (30) years ago and the records of the packaging process and configuration were not documented as waste. Therefore, it seemed that the load plans were prepared with limited information regarding parent container configuration, hence in the dark. The design of the Auxiliary Hot Cell Facility (AHCF) in technical area (TA) V, where the repackaging was to take place, along with the shipping requirements for the RH 72-B cask, limited the options for repackaging. This configuration required that 0.114 m{sup 3} (30-gallon) containers be packaged in the hot cell, removed and placed into 0.208 m{sup 3} (55-gallon) containers to mitigate contamination issues. The material had been tracked as accountable nuclear material and Sandia had verified Material Control and Accountability (MC and A) data on each parent container. Based on the MC and A data, scaling factors could be determined and sampling and analysis was not required. However, it meant that parent containers could not be split into two daughter containers, though two or more parent containers could be combined into one daughter. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Reliability Assessment of Remote Visual Examination

Remote visual testing (RVT) is a commonly used nondestructive examination method for inservice inspection (ISI) of reactor internals to detect cracking and gross component failures. Despite widespread use, the detection reliability of RVT and the factors that impact overall RVT performance have been unresolved issues. This report describes the results from an assessment sponsored by the U.S. Nuclear Regulatory Commission and conducted by the Pacific Northwest National Laboratory, in cooperation with the Electric Power Research Institute, for evaluating the reliability of RVT methods currently being used for [reactor] in-vessel visual inspection. The goal was to assess the performance of commercially applied RVT examination procedures implemented by qualified personnel, as well as to identify and qualitatively assess enhancements to examination procedures for detecting flaws in test specimens. The assessment was performed over three phases of research, with the results of the first phase used to identify key controllable parameters impacting RVT performance and to design the next two phases (round-robin testing). Participants in the round-robin tests included teams of inspectors from commercial nuclear power ISI vendors. Participants were asked to determine if a specimen contained a crack and its approximate location, orientation, and length. The detection and location information were used to compute estimates of the probability of detection and false call probability for various scenarios such as the placement and orientation of the flaw on the specimen, team-to-team variations in detection performance, and effects of secondary review on RVT detection. Results showed that crack opening displacement (COD) is the dominant factor in the reliability of crack detection using commercially applied RVT procedures, with crack length being weakly correlated with detection probability. The implication is that RVT detection is likely heavily dependent on the contrast produced by the crack opening, with crack detection becoming less reliable as the COD decreases. Further, the results indicated RVT will be challenged when cracks are located in the vicinity of surface features such as scratches or weld ripples, or close to the edge of welds where shadowing and/or the presence of weld undercuts may complicate the ability to detect the crack. The assessment also reinforced earlier findings on the importance of lighting in flaw detection and appeared to align with other studies that find improved reliability when using multiple inspectors or analysts. Finally, the results pointed to the importance of practice with specimens that mimic conditions likely to be found in the field. Based on the findings and limitations of the assessment, a number of recommendations are made regarding best practices for improving the reliability of RVT in a field setting and for addressing remaining informational gaps. This report describes the design and noted limitations of the three phases of research, the analysis methodology for each phase, and the results of the research. The results of this assessment provide a benchmark set of data on the reliability of RVT for detecting cracking, assuming the implementation of field-like inspection procedures. The likely impact of several uncontrolled factors on RVT detection performance are discussed, and recommendations regarding the use of these results to assess field performance are provided. Finally, recent advances in RVT technology are briefly discussed and point to the potential need for continued research to evaluate the capability and effectiveness of the technique as improvements are implemented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Irradiation-induced formation of a spinel phase at the FeCr/MgO interface

Oxide dispersion strengthened ferritic/martensitic alloys have attracted significant attention for their potential uses in future nuclear reactors and storage vessels, as the metal/oxide interfaces act as stable high-strength sinks for point defects while also dispersing helium. Here, in order to unravel the evolution and interplay of interface structure and chemistry upon irradiation in these types of materials, an atomically sharp FeCr/MgO interface was synthesized at 500 °C and separately annealed and irradiated with Ni 3+ ions at 500 °C. After annealing, a slight enrichment of Cr atoms was observed at the interface, but no other structural changes were found. However, under irradiation, sufficient Cr diffuses across the interface into the MgO to form a Cr-enriched transition layer that contains spinel precipitates. First-principles calculations indicate that it is energetically favorable to incorporate Cr, but not Fe, substitutionally into MgO. Furthermore, our results indicate that irradiation can be used to form new phases and complexions at interfaces, which may have different radiation tolerance than the pristine structures.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Modeling an unmitigated thermal quench event in a large field magnet in a DEMO reactor

The superconducting magnet systems of future fusion reactors, such as a Demonstration Power Plant (DEMO), will produce magnetic field energies in the 10 s of GJ range. The release of this energy during a fault condition could produce arcs that can damage the magnets of these systems. The public safety consequences of such events must be explored for a DEMO reactor because the magnets are located near the DEMO's primary radioactive confinement barrier, the reactor's vacuum vessel (VV). Great care will be taken in the design of DEMO's magnet systems to detect and provide a rapid field energy dump to avoid any accidents conditions. During an event when a fault condition proceeds undetected, the potential of producing melting of the magnet exists. If molten material from the magnet impinges on the walls of the VV, these walls could fail, resulting in a pathway for release of radioactive material from the VV. A model is under development at Idaho National Laboratory (INL) called MAGARC to investigate the consequences of this accident in a large toroidal field (TF) coil. Recent improvements to this model are described in this paper, along with predictions for a DEMO relevant event in a toroidal field magnet.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Behavior of the Cadmium Pool in the INL Mk-IV Electrorefiner

The Mk-IV electrorefiner has been in service since 1996 recovering uranium from spent fuels from the Experimental Breeder Rector II and Fast Flux Test Facility sodium-cooled fast reactors. The electrorefiner vessel includes a cadmium pool beneath the salt. The voltage signal between the cadmium pool and reference electrode (Ag/AgCl type) provides information about the chemical condition of the cadmium pool with respect to uranium and zirconium saturation. This information is used to guide process control and sampling decisions. This paper describes how the voltage signal is related to chemistry by the analyses of data from Mk-IV operations and confirmatory experimentation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Harvesting Energy from Wastewater by Converting Sewage

This project aims were to develop and demonstrate a scalable, integrated process to convert sewage sludge into renewable natural gas (RNG), enabling wastewater treatment plants (WWTPs) to become net energy producers. The system proposal integrates autothermal hydrothermal liquefaction (AT-HTL), supercritical salt precipitation (SCSP), and hydrothermal gasification (HTG), collectively forming the Supercritical Sludge-to-Gas (SC-S2G) platform. Initially, batch hydrothermal liquefaction reactions were used to screen sewage sludge using AT-HTL (later termed RI-HTL) conversion to biocrude, aqueous and char phases compared to hydrothermal liquefaction (HTL). Significant improvement in biocrude yield using peroxide addition at O:C ratio of 0.05 and under conditions of 300°C for 10 minutes gave 57% biocrude yield and 85% fluid carbon yield (biocrude plus aqueous), while minimizing the loss of carbon to char solids (~7%). Hence, RI-HTL was shown to be effective for conversion of real sewage sludge. The corrosion of the alloy reactor tubes or vessels is an important factor when developing a process that includes an oxidant and a chemically complex feed like sewage sludge. We investigated the corrosion rates on metal alloys at 350°C for 240 hours. Corrosion rates of 0.21 and 0.26 mpy for 304L and 316L stainless steel were measured respectively. The corrosion information obtained in this investigation was utilized by PNNL for design, materials sourcing and construction of the pilot scale continuous flow system.

09 BIOMASS FUELS↗
NASA NTRS1971

The open-cycle gas-core nuclear rocket engine - Some engineering considerations.

A preliminary design study of a conceptual 6000-MW open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 44,200 lb and a specific impulse of 4400 sec. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel) and the waste heat rejection system were considered conceptually and were sized.

Taylor, M. F.↗
DOE OSTI2016

Packed Bed thermal storage for compact Light water-cooled Small Modular Reactors

Nuclear power plants (NPPs) have less flexibility to follow grid load demand than their natural-gas fueled counterparts. Most of the natural-gas power plants can supply peak-loads by adding more fuel, and thus, can generate far more revenue during those peak hours. On the other hand, use of NPPs for peak load following is quite complex due to technical constraints associated with reactor behavior. These technical challenges include the adequate handling of reactivity swings caused by time-varying fuel and moderator temperatures, a higher fuel-failure probability due to thermal-structural cycling, and spatial variations in xenon concentration. Although there are presently some reactors around the world that are operating with flexible load-following capabilities, such operation is restricted to slowly-varying powers, 2-3 times a day, and only up to 80% of the fuel cycle. Thus, a more convenient and effective method to facilitate load following by NPPs would be to integrate energy storage Authors have previously developed models and performed exergy analysis for integrating liquid type TES with LWRs and SMRs. These exergy based models show that technical feasibility index of integrating thermal energy storage is very high. But most of these previous studies include integration based on existing layout of NPPs with heat exchange between storage Heat Transfer Fluid (HTF) and Reactor Coolant (RC). However, most of the Light Water-cooled Small modular reactor (LW-SMR) based NPP designs can not accommodate additional heat exchange equipment. Although power production is only reduced to 1/20 scale, the containment volume in thus NuScale LW-SMR is reduced to {approx} 1/170 scale. As steam is the secondary or working fluid which exits out of the reactor containment in LW-SMRs, the simplest and most practical way to integrate thermal storage with LW-SMRs is to directly store the energy from steam. Indirect heat exchange between steam and another heat transfer fluid during the storage process, and then heat recovery again via indirect heat exchange is inefficient process due to heat transfer resistance. This inefficiency is due to what is called as 'Pinch point' which occurs whenever there is indirect heat transfer between two media where one undergoes phase change and other undergoes change in sensible heat. Another existing popular option is to use a steam accumulator which stores dry or wet steam directly inside the pressure vessels, however, due to large volumetric requirements it is not a viable option. This work proposes a new method of integrating packed bed TES systems with LW-SMRs, where steam produced in the secondary side of LW-SMRs can be injected into the packed bed of inert particles with sufficient thermal conductivity and capacity to efficiently store energy with high energy density. The changes in the load demand will not effect the LWSMR operation and the excess steam will be diverted to TES. When the demand increases this stored energy in hot packed beds can be recovered by injecting cold pressurized water. Although there are some existing packed bed TES designs and models which consider steam as HTFs, these models have not been extended for wide range of process parameters important for this integration. Authors have previously developed a packed bed thermal model for steam injection and validated it using experimental data. This thermal model is extended to assess behavior of packed bed thermal energy storage (TES) system under storage and recovery cycles. Steam is the only heat transfer fluid (HTF) which exits the reactor vessel, so any application such as process heat or TES integration should be evaluated with steam as process fluid. This project will develop a process system integration model of LW-SMR with packed bed type TES using steam as HTF. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Detecting Thermally Induced Spinodal Decomposition with Picosecond Ultrasonics in Cast Austenitic Stainless Steels

Given the existential climate crisis faced by mankind and the world, the lifetime and sustainability of nuclear reactors as a carbon-free source of renewable energy depend on the susceptibility of their structural components to environmental degradation. In particular, critical components for light water reactors (LWRs) evolve over decades in service, losing ductility and toughness due to thermal and irradiation aging. Techniques to monitor their health cannot be easily applied in the field due to their destructive, expensive, or immobile nature. Thus, non-destructive evaluation (NDE) methods are sought to monitor and evaluate the health of major LWR components such as core barrels, steam generator tubes, or primary coolant pipes and are often required by policy, such as NRC policy #10-CFR-50.65. In this work we demonstrate the use of gigahertz, non-contact ultrasonics to gauge the state of cast austenitic stainless steels (CASS), used in some of the largest components in LWR primary systems. We do so by linking changes in their surface acoustic wave (SAW) characteristics using transient grating spectroscopy (TGS) to transmission electron microscopy (TEM)-verified evidence of spinodal decomposition and G-phase precipitation. In this thesis, thermal aging is shown to induce SAW peak splitting in spinodally decomposed CASS alloys, correlated strongly with lowered toughness and decreased ductility. Furthermore, statistical testing on the number of SAW peak splits observed show that the second SAW peak significantly appears more frequently and is significantly different in frequency in comparison to counts and frequencies measured in unaged specimens. The ability of this technique to non-destructively detect microstructural degradation at a distance in a predictive manner in the case of CASS motivates extending gigahertz ultrasonics to detect other LWR material degradation modes as an in-vessel inspection technique, such as reactor pressure vessel (RPV) embrittlement. This allows for the greater use of NDE techniques for confident monitoring of LWR structural material health to 80 years and beyond, saving costs by minimizing structural replacements until needed and maximizing energy production by preventing early decommission until necessary.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Nuclear reactor support and seismic restraint with core retention cooling features

A nuclear island includes a nuclear reactor, a lateral seismic restraint, and a reactor core retention cooling system. The lateral seismic restraint includes a vertically oriented pin attached to one of the bottom of the lower vessel head and the floor underneath the nuclear reactor, and a mating pin socket is attached to the other of the bottom of the lower vessel head and the floor. The reactor core retention cooling system includes one or more baffles, optionally thermally insulating material, disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head. A plenum is defined between the one or more baffles and the exterior surface of a lower portion of the reactor pressure vessel. The one or more baffles may define a lower plenum inlet surrounding the lateral seismic restraint.

Edwards, Tyler A.↗