Geoantineutrino spectrum and slow nuclear burning on the boundary of the liquid and solid phases of the Earth's core

NASA Astrophysics Data System (ADS)

Rusov, V. D.; Pavlovich, V. N.; Vaschenko, V. N.; Tarasov, V. A.; Zelentsova, T. N.; Bolshakov, V. N.; Litvinov, D. A.; Kosenko, S. I.; Byegunova, O. A.

2007-09-01

We give an alternative description of the data produced in the KamLAND experiment. Assuming the existence of a natural nuclear reactor on the boundary of the liquid and solid phases of the Earth's core, a geoantineutrino spectrum is obtained. This assumption is based on the experimental results of V. Anisichkin and his collaborators on the interaction of uranium dioxide and uranium carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600-2200°C), which led to the proposal of the existence of an actinide shell in the Earth's core. We describe the operating mechanism of this reactor as solitary waves of nuclear burning in 238U and/or 232Th medium, in particular, as neutron fission progressive waves of Feoktistov and/or Teller et al. type. Next, we propose a simplified model for the accumulation and burn-up kinetics in Feoktistov's U-Pu fuel cycle. We also apply this model for numerical simulations of neutron fission wave in a two-phase UO2/Fe medium on the surface of the Earth's solid core. The proposed georeactor model offers a mechanism for the generation of 3He. The 3He/4He distribution in the Earth's interior is calculated, which in turn can be used as a natural quantitative criterion of the georeactor thermal power. Finally, we give a tentative estimation of the geoantineutrino intensity and spectrum on the Earth's surface. For this purpose we use the O'Nions et al. geochemical model of mantle differentiation and crust growth complemented by a nuclear energy source (georeactor with power of 30 TW).

Multiphysics Thermal-Fluid Design Analysis of a Non-Nuclear Tester for Hot-Hydrogen Materials and Component Development

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Foote, John; Litchford, Ron

2006-01-01

The objective of this effort is to perform design analyses for a non-nuclear hot-hydrogen materials tester, as a first step towards developing efficient and accurate multiphysics, thermo-fluid computational methodology to predict environments for hypothetical solid-core, nuclear thermal engine thrust chamber design and analysis. The computational methodology is based on a multidimensional, finite-volume, turbulent, chemically reacting, thermally radiating, unstructured-grid, and pressure-based formulation. The multiphysics invoked in this study include hydrogen dissociation kinetics and thermodynamics, turbulent flow, convective, and thermal radiative heat transfers. The goals of the design analyses are to maintain maximum hot-hydrogen jet impingement energy and to minimize chamber wall heating. The results of analyses on three test fixture configurations and the rationale for final selection are presented. The interrogation of physics revealed that reactions of hydrogen dissociation and recombination are highly correlated with local temperature and are necessary for accurate prediction of the hot-hydrogen jet temperature.

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, Vladimir

1986-01-01

This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

Advanced nuclear thermal propulsion concepts

NASA Technical Reports Server (NTRS)

Howe, Steven D.

1993-01-01

In 1989, a Presidential directive created the Space Exploration Initiative (SEI) which had a goal of placing mankind on Mars in the early 21st century. The SEI was effectively terminated in 1992 with the election of a new administration. Although the initiative did not exist long enough to allow substantial technology development, it did provide a venue, for the first time in 20 years, to comprehensively evaluate advanced propulsion concepts which could enable fast, manned transits to Mars. As part of the SEI based investigations, scientists from NASA, DoE National Laboratories, universities, and industry met regularly and proceeded to examine a variety of innovative ideas. Most of the effort was directed toward developing a solid-core, nuclear thermal rocket and examining a high-power nuclear electric propulsion system. In addition, however, an Innovative Concepts committee was formed and charged with evaluating concepts that offered a much higher performance but were less technologically mature. The committee considered several concepts and eventually recommended that further work be performed in the areas of gas core fission rockets, inertial confinement fusion systems, antimatter based rockets, and gas core fission electric systems. Following the committee's recommendations, some computational modeling work has been performed at Los Alamos in certain of these areas and critical issues have been identified.

Direct measurement of thermal conductivity in solid iron at planetary core conditions.

PubMed

Konôpková, Zuzana; McWilliams, R Stewart; Gómez-Pérez, Natalia; Goncharov, Alexander F

2016-06-02

The conduction of heat through minerals and melts at extreme pressures and temperatures is of central importance to the evolution and dynamics of planets. In the cooling Earth's core, the thermal conductivity of iron alloys defines the adiabatic heat flux and therefore the thermal and compositional energy available to support the production of Earth's magnetic field via dynamo action. Attempts to describe thermal transport in Earth's core have been problematic, with predictions of high thermal conductivity at odds with traditional geophysical models and direct evidence for a primordial magnetic field in the rock record. Measurements of core heat transport are needed to resolve this difference. Here we present direct measurements of the thermal conductivity of solid iron at pressure and temperature conditions relevant to the cores of Mercury-sized to Earth-sized planets, using a dynamically laser-heated diamond-anvil cell. Our measurements place the thermal conductivity of Earth's core near the low end of previous estimates, at 18-44 watts per metre per kelvin. The result is in agreement with palaeomagnetic measurements indicating that Earth's geodynamo has persisted since the beginning of Earth's history, and allows for a solid inner core as old as the dynamo.

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ehud Greenspan

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Modeling of Thermal Phase Noise in a Solid Core Photonic Crystal Fiber-Optic Gyroscope.

PubMed

Song, Ningfang; Ma, Kun; Jin, Jing; Teng, Fei; Cai, Wei

2017-10-26

A theoretical model of the thermal phase noise in a square-wave modulated solid core photonic crystal fiber-optic gyroscope has been established, and then verified by measurements. The results demonstrate a good agreement between theory and experiment. The contribution of the thermal phase noise to the random walk coefficient of the gyroscope is derived. A fiber coil with 2.8 km length is used in the experimental solid core photonic crystal fiber-optic gyroscope, showing a random walk coefficient of 9.25 × 10 -5 deg/√h.

Chemical and thermal stability of core-shelled magnetite nanoparticles and solid silica

NASA Astrophysics Data System (ADS)

Cendrowski, Krzysztof; Sikora, Pawel; Zielinska, Beata; Horszczaruk, Elzbieta; Mijowska, Ewa

2017-06-01

Pristine nanoparticles of magnetite were coated by solid silica shell forming core/shell structure. 20 nm thick silica coating significantly enhanced the chemical and thermal stability of the iron oxide. Chemical and thermal stability of this structure has been compared to the magnetite coated by mesoporous shell and pristine magnetite nanoparticles. It is assumed that six-membered silica rings in a solid silica shell limit the rate of oxygen diffusion during thermal treatment in air and prevent the access of HCl molecules to the core during chemical etching. Therefore, the core/shell structure with a solid shell requires a longer time to induce the oxidation of iron oxide to a higher oxidation state and, basically, even strong concentrated acid such as HCl is not able to dissolve it totally in one month. This leads to the desired performance of the material in potential applications such as catalysis and environmental protection.

Modeling of Thermal Phase Noise in a Solid Core Photonic Crystal Fiber-Optic Gyroscope

PubMed Central

Song, Ningfang; Ma, Kun; Jin, Jing; Teng, Fei; Cai, Wei

2017-01-01

A theoretical model of the thermal phase noise in a square-wave modulated solid core photonic crystal fiber-optic gyroscope has been established, and then verified by measurements. The results demonstrate a good agreement between theory and experiment. The contribution of the thermal phase noise to the random walk coefficient of the gyroscope is derived. A fiber coil with 2.8 km length is used in the experimental solid core photonic crystal fiber-optic gyroscope, showing a random walk coefficient of 9.25 × 10−5 deg/h. PMID:29072605

Fission fragment assisted reactor concept for space propulsion: Foil reactor

NASA Technical Reports Server (NTRS)

Wright, Steven A.

1991-01-01

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

Gas-core reactor power transient analysis

NASA Technical Reports Server (NTRS)

Kascak, A. F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

Free-Standing and Self-Crosslinkable Hybrid Films by Core-Shell Particle Design and Processing.

PubMed

Vowinkel, Steffen; Paul, Stephen; Gutmann, Torsten; Gallei, Markus

2017-11-15

The utilization and preparation of functional hybrid films for optical sensing applications and membranes is of utmost importance. In this work, we report the convenient and scalable preparation of self-crosslinking particle-based films derived by directed self-assembly of alkoxysilane-based cross-linkers as part of a core-shell particle architecture. The synthesis of well-designed monodisperse core-shell particles by emulsion polymerization is the basic prerequisite for subsequent particle processing via the melt-shear organization technique. In more detail, the core particles consist of polystyrene (PS) or poly(methyl methacrylate) (PMMA), while the comparably soft particle shell consists of poly(ethyl acrylate) (PEA) and different alkoxysilane-based poly(methacrylate)s. For hybrid film formation and convenient self-cross-linking, different alkyl groups at the siloxane moieties were investigated in detail by solid-state Magic-Angle Spinning Nuclear Magnetic Resonance (MAS, NMR) spectroscopy revealing different crosslinking capabilities, which strongly influence the properties of the core or shell particle films with respect to transparency and iridescent reflection colors. Furthermore, solid-state NMR spectroscopy and investigation of the thermal properties by differential scanning calorimetry (DSC) measurements allow for insights into the cross-linking capabilities prior to and after synthesis, as well as after the thermally and pressure-induced processing steps. Subsequently, free-standing and self-crosslinked particle-based films featuring excellent particle order are obtained by application of the melt-shear organization technique, as shown by microscopy (TEM, SEM).

Horizontal baffle for nuclear reactors

DOEpatents

Rylatt, John A.

1978-01-01

A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

Thermal Evolution of Earht's Core during Accretion: a Preliminary Solid Inner Core at the End of Accrfetion.

NASA Astrophysics Data System (ADS)

Arkani-Hamed, J.

2015-12-01

Growth of an inner core has conventionally been related to core cooling blow the liquidus of iron. It is however possible that the core of the proto-Earth solidifies upon pressure increase during accretion. The lithostatic pressure in the proto-Earth increases immediately after merging each impactor, and the pressure-dependent liquidus of iron may supersede the temperature near the center resulting in a solid inner core. Assuming that Earth is formed by accreting a few dozen Moon to Mars size planetary embryos, the thermal evolution of the proto-Earth's core is investigated during accretion. The collision of an embryo heats the Earth differentially and the rotating low-viscosity, differentially heated core stratifies, creating a spherically symmetric stable and radially increasing temperature distribution. Convection occurs in the outer core while heat transfers by conduction in deeper parts. It is assumed that the iron core of an embryo pools at the bottom of partially molten mantle and thermally equilibrates with surroundings. It then descends as an iron diapir in the solid silicate mantle, while releasing its gravitational energy. Depending on its temperature when arrives at the core mantle boundary, it may spread on the core creating a hot layer or plunge into the core and descend to a neutrally buoyant level while further releasing its gravitational energy. A few dozen thermal evolution models of the core are investigates to examine effects of major parameters such as: total number of impacting embryos; partitioning of the gravitational energy released during the descent of the diaper in the mantle (between the silicate mantle and the iron diaper), and in the core (between the proto-Earth's core and that of the embryo); and gravitational energy and latent heat released due to the core solidification. All of the models predict a large solid inner core, about 1500 to 2000 km in radius, at the end of accretion.

Multiphysics Nuclear Thermal Rocket Thrust Chamber Analysis

NASA Technical Reports Server (NTRS)

Wang, Ten-See

2005-01-01

The objective of this effort is t o develop an efficient and accurate thermo-fluid computational methodology to predict environments for hypothetical thrust chamber design and analysis. The current task scope is to perform multidimensional, multiphysics analysis of thrust performance and heat transfer analysis for a hypothetical solid-core, nuclear thermal engine including thrust chamber and nozzle. The multiphysics aspects of the model include: real fluid dynamics, chemical reactivity, turbulent flow, and conjugate heat transfer. The model will be designed to identify thermal, fluid, and hydrogen environments in all flow paths and materials. This model would then be used to perform non- nuclear reproduction of the flow element failures demonstrated in the Rover/NERVA testing, investigate performance of specific configurations and assess potential issues and enhancements. A two-pronged approach will be employed in this effort: a detailed analysis of a multi-channel, flow-element, and global modeling of the entire thrust chamber assembly with a porosity modeling technique. It is expected that the detailed analysis of a single flow element would provide detailed fluid, thermal, and hydrogen environments for stress analysis, while the global thrust chamber assembly analysis would promote understanding of the effects of hydrogen dissociation and heat transfer on thrust performance. These modeling activities will be validated as much as possible by testing performed by other related efforts.

Parameter study of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239c

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, W.W.; Layton, J.P.

1976-09-13

The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The NUROC3A systems analysis code was designed to provide the user with performance characteristics of the dual-mode system. Volume 3 describes utilization of the NUROC3A code to produce a detailed parameter study of the system.

Thermal margin protection system for a nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Musick, C.R.

1974-02-12

A thermal margin protection system for a nuclear reactor is described where the coolant flow flow trip point and the calculated thermal margin trip point are switched simultaneously and the thermal limit locus is made more restrictive as the allowable flow rate is decreased. The invention is characterized by calculation of the thermal limit Locus in response to applied signals which accurately represent reactor cold leg temperature and core power; cold leg temperature being corrected for stratification before being utilized and reactor power signals commensurate with power as a function of measured neutron flux and thermal energy added to themore » coolant being auctioneered to select the more conservative measure of power. The invention further comprises the compensation of the selected core power signal for the effects of core radial peaking factor under maximum coolant flow conditions. (Official Oazette)« less

Prediction of Ablation Rates from Solid Surfaces Exposed to High Temperature Gas Flow

NASA Technical Reports Server (NTRS)

Akyuzlu, Kazim M.; Coote, David

2013-01-01

A mathematical model and a solution algorithm is developed to study the physics of high temperature heat transfer and material ablation and identify the problems associated with the flow of hydrogen gas at very high temperatures and velocities through pipes and various components of Nuclear Thermal Rocket (NTR) motors. Ablation and melting can be experienced when the inner solid surface of the cooling channels and the diverging-converging nozzle of a Nuclear Thermal Rocket (NTR) motor is exposed to hydrogen gas flow at temperatures around 2500 degrees Kelvin and pressures around 3.4 MPa. In the experiments conducted on typical NTR motors developed in 1960s, degradation of the cooling channel material (cracking in the nuclear fuel element cladding) and in some instances melting of the core was observed. This paper presents the results of a preliminary study based on two types of physics based mathematical models that were developed to simulate the thermal-hydrodynamic conditions that lead to ablation of the solid surface of a stainless steel pipe exposed to high temperature hydrogen gas near sonic velocities. One of the proposed models is one-dimensional and assumes the gas flow to be unsteady, compressible and viscous. An in-house computer code was developed to solve the conservations equations of this model using a second-order accurate finite-difference technique. The second model assumes the flow to be three-dimensional, unsteady, compressible and viscous. A commercial CFD code (Fluent) was used to solve the later model equations. Both models assume the thermodynamic and transport properties of the hydrogen gas to be temperature dependent. In the solution algorithm developed for this study, the unsteady temperature of the pipe is determined from the heat equation for the solid. The solid-gas interface temperature is determined from an energy balance at the interface which includes heat transfer from or to the interface by conduction, convection, radiation, and ablation. Two different ablation models are proposed to determine the heat loss from the solid surface due to the ablation of the solid material. Both of them are physics based. Various numerical simulations were carried out using both models to predict the temperature distribution in the solid and in the gas flow, and then predict the ablation rates at a typical NTR motor hydrogen gas temperature and pressure. Solid mass loss rate per foot of a pipe was also calculated from these predictions. The results are presented for fully developed turbulent flow conditions in a sample SS pipe with a 6 inch diameter.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-03-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Emergency deployable core catcher

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosewell, M.P.

An emergency melt down core catcher apparatus for a nuclear reactor having a retrofitable eutectic solute holding vessel connected to a core containment vessel with particle transferring fluid and particles or granules of solid eutectic solute materials contained therein and transferable by automatically operated valve means to transport and position the solid eutectic solute material in a position below the core to catch and react with any partial or complete melt down of the fuel core.

Thermal Dispersion Within a Porous Medium Near a Solid Wall

NASA Technical Reports Server (NTRS)

Simon, T.; McFadden, G.; Ibrahim, M.

2006-01-01

The regenerator is a key component to Stirling cycle machine efficiency. Typical regenerators are of sintered fine wires or layers of fine-wire screens. Such porous materials are contained within solid-waH casings. Thermal energy exchange between the regenerator and the casing is important to cycle performance for the matrix and casing would not have the same axial temperature profile in an actual machine. Exchange from one to the other may allow shunting of thermal energy, reducing cycle efficiency. In this paper, temperature profiles within the near-wall region of the matrix are measured and thermal energy transport, termed thermal dispersion, is inferred. The data show how the wall affects thermal transport. Transport normal to the mean flow direction is by conduction within the solid and fluid and by advective transport within the matrix. In the near-wall region, both may be interrupted from their normal in-core pattern. Solid conduction paths are broken and scales of advective transport are damped. An equation is presented which describes this change for a wire screen mesh. The near-wall layer typically acts as an insulating layer. This should be considered in design or analysis. Effective thermal conductivity within the core is uniform. In-core transverse thermal effective conductivity values are compared to direct and indirect measurements reported elsewhere and to 3D numerical simulation results, computed previously and reported elsewhere. The 3-D CFD model is composed of six cylinders in cross flow, staggered in arrangement to match the dimensions and porosity of the matrix used in the experiments. The commercial code FLUENT is used to obtain the flow and thermal fields. The thermal dispersion and effective thermal conductivities for the matrix are computed from the results.

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

Non-invasive, transient determination of the core temperature of a heat-generating solid body

PubMed Central

Anthony, Dean; Sarkar, Daipayan; Jain, Ankur

2016-01-01

While temperature on the surface of a heat-generating solid body can be easily measured using a variety of methods, very few techniques exist for non-invasively measuring the temperature inside the solid body as a function of time. Measurement of internal temperature is very desirable since measurement of just the surface temperature gives no indication of temperature inside the body, and system performance and safety is governed primarily by the highest temperature, encountered usually at the core of the body. This paper presents a technique to non-invasively determine the internal temperature based on the theoretical relationship between the core temperature and surface temperature distribution on the outside of a heat-generating solid body as functions of time. Experiments using infrared thermography of the outside surface of a thermal test cell in a variety of heating and cooling conditions demonstrate good agreement of the predicted core temperature as a function of time with actual core temperature measurement using an embedded thermocouple. This paper demonstrates a capability to thermally probe inside solid bodies in a non-invasive fashion. This directly benefits the accurate performance prediction and control of a variety of engineering systems where the time-varying core temperature plays a key role. PMID:27804981

Non-invasive, transient determination of the core temperature of a heat-generating solid body

NASA Astrophysics Data System (ADS)

Anthony, Dean; Sarkar, Daipayan; Jain, Ankur

2016-11-01

While temperature on the surface of a heat-generating solid body can be easily measured using a variety of methods, very few techniques exist for non-invasively measuring the temperature inside the solid body as a function of time. Measurement of internal temperature is very desirable since measurement of just the surface temperature gives no indication of temperature inside the body, and system performance and safety is governed primarily by the highest temperature, encountered usually at the core of the body. This paper presents a technique to non-invasively determine the internal temperature based on the theoretical relationship between the core temperature and surface temperature distribution on the outside of a heat-generating solid body as functions of time. Experiments using infrared thermography of the outside surface of a thermal test cell in a variety of heating and cooling conditions demonstrate good agreement of the predicted core temperature as a function of time with actual core temperature measurement using an embedded thermocouple. This paper demonstrates a capability to thermally probe inside solid bodies in a non-invasive fashion. This directly benefits the accurate performance prediction and control of a variety of engineering systems where the time-varying core temperature plays a key role.

Thermal Equation of State of Iron: Constraint on the Density Deficit of Earth's Core

NASA Astrophysics Data System (ADS)

Fei, Y.; Murphy, C. A.; Shibazaki, Y.; Huang, H.

2013-12-01

The seismically inferred densities of Earth's solid inner core and the liquid outer core are smaller than the measured densities of solid hcp-iron and liquid iron, respectively. The inner core density deficit is significantly smaller than the outer core density deficit, implying different amounts and/or identities of light-elements incorporated in the inner and outer cores. Accurate measurements of the thermal equation-of-state of iron over a wide pressure and temperature range are required to precisely quantify the core density deficits, which are essential for developing a quantitative composition model for the core. The challenge has been evaluating the experimental uncertainties related to the choice of pressure scales and the sample environment, such as hydrostaticity at multi-megabar pressures and extreme temperatures. We have conducted high-pressure experiments on iron in MgO, NaCl, and Ne pressure media and obtained in-situ X-ray diffraction data up to 200 GPa at room temperature. Using inter-calibrated pressure scales including the MgO, NaCl, Ne, and Pt scales, we have produced a consistent compression curve of hcp-Fe at room temperature. We have also performed laser-heated diamond-anvil cell experiments on both Fe and Pt in a Ne pressure medium. The experiment was designed to quantitatively compare the thermal expansion of Fe and Pt in the same sample environment using Ne as the pressure medium. The thermal expansion data of hcp-Fe at high pressure were derived based on the thermal equation of state of Pt. Using the 300-K isothermal compression curve of iron derived from our static experiments as a constraint, we have developed a thermal equation of state of hcp-Fe that is consistent with the static P-V-T data of iron and also reproduces the shock wave Hugoniot data for pure iron. The thermodynamic model, based on both static and dynamic data, is further used to calculate the density and bulk sound velocity of liquid iron. Our results define the solid inner core and liquid outer core density deficits, which can serve as the basis for any core composition models.

A New Capability for Nuclear Thermal Propulsion Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

Open cycle gas core nuclear rockets

NASA Technical Reports Server (NTRS)

Ragsdale, Robert

1991-01-01

The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.

The Liquid Annular Reactor System (LARS) propulsion

NASA Technical Reports Server (NTRS)

Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

1990-01-01

A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

Palaeointensity, core thermal conductivity and the unknown age of the inner core

NASA Astrophysics Data System (ADS)

Smirnov, Aleksey V.; Tarduno, John A.; Kulakov, Evgeniy V.; McEnroe, Suzanne A.; Bono, Richard K.

2016-05-01

Data on the evolution of Earth's magnetic field intensity are important for understanding the geodynamo and planetary evolution. However, the paleomagnetic record in rocks may be adversely affected by many physical processes, which must be taken into account when analysing the palaeointensity database. This is especially important in the light of an ongoing debate regarding core thermal conductivity values, and how these relate to the Precambrian geodynamo. Here, we demonstrate that several data sets in the Precambrian palaeointensity database overestimate the true paleofield strength due to the presence of non-ideal carriers of palaeointensity signals and/or viscous re-magnetizations. When the palaeointensity overestimates are removed, the Precambrian database does not indicate a robust change in geomagnetic field intensity during the Mesoproterozoic. These findings call into question the recent claim that the solid inner core formed in the Mesoproterozoic, hence constraining the thermal conductivity in the core to `moderate' values. Instead, our analyses indicate that the presently available palaeointensity data are insufficient in number and quality to constrain the timing of solid inner core formation, or the outstanding problem of core thermal conductivity. Very young or very old inner core ages (and attendant high or low core thermal conductivity values) are consistent with the presently known history of Earth's field strength. More promising available data sets that reflect long-term core structure are geomagnetic reversal rate and field morphology. The latter suggests changes that may reflect differences in Archean to Proterozoic core stratification, whereas the former suggest an interval of geodynamo hyperactivity at ca. 550 Ma.

Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

NASA Astrophysics Data System (ADS)

Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

2004-02-01

A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

Radio-Frequency Driven Dielectric Heaters for Non-Nuclear Testing in Nuclear Core Development

NASA Technical Reports Server (NTRS)

Sims, William Herbert, III (Inventor); Godfroy, Thomas J. (Inventor); Bitteker, Leo (Inventor)

2006-01-01

Apparatus and methods are provided through which a radiofrequency dielectric heater has a cylindrical form factor, a variable thermal energy deposition through variations in geometry and composition of a dielectric, and/or has a thermally isolated power input.

Atmospheric Mining in the Outer Solar System: Resource Capturing, Storage, and Utilization

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2014-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as helium 3 and hydrogen can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and hydrogen (deuterium, etc.) were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate for hydrogen helium 4 and helium 3, storage options, and different methods of direct use of the captured gases. Additional supporting analyses were conducted to illuminate vehicle sizing and orbital transportation issues.

Solar System Exploration Augmented by Lunar and Outer Planet Resource Utilization: Historical Perspectives and Future Possibilities

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2014-01-01

Establishing a lunar presence and creating an industrial capability on the Moon may lead to important new discoveries for all of human kind. Historical studies of lunar exploration, in-situ resource utilization (ISRU) and industrialization all point to the vast resources on the Moon and its links to future human and robotic exploration. In the historical work, a broad range of technological innovations are described and analyzed. These studies depict program planning for future human missions throughout the solar system, lunar launched nuclear rockets, and future human settlements on the Moon, respectively. Updated analyses based on the visions presented are presented. While advanced propulsion systems were proposed in these historical studies, further investigation of nuclear options using high power nuclear thermal propulsion, nuclear surface power, as well as advanced chemical propulsion can significantly enhance these scenarios. Robotic and human outer planet exploration options are described in many detailed and extensive studies. Nuclear propulsion options for fast trips to the outer planets are discussed. To refuel such vehicles, atmospheric mining in the outer solar system has also been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and hydrogen can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and hydrogen (deuterium, etc.) were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses have investigated resource capturing aspects of atmospheric mining in the outer solar system. These analyses included the gas capturing rate, storage options, and different methods of direct use of the captured gases. While capturing 3He, large amounts of hydrogen and 4He are produced. With these two additional gases, the potential for fueling small and large fleets of additional exploration and exploitation vehicles exists.

Solar System Exploration Augmented by Lunar and Outer Planet Resource Utilization: Historical Perspectives and Future Possibilities

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2014-01-01

Establishing a lunar presence and creating an industrial capability on the Moon may lead to important new discoveries for all of human kind. Historical studies of lunar exploration, in-situ resource utilization (ISRU) and industrialization all point to the vast resources on the Moon and its links to future human and robotic exploration. In the historical work, a broad range of technological innovations are described and analyzed. These studies depict program planning for future human missions throughout the solar system, lunar launched nuclear rockets, and future human settlements on the Moon, respectively. Updated analyses based on the visions presented are presented. While advanced propulsion systems were proposed in these historical studies, further investigation of nuclear options using high power nuclear thermal propulsion, nuclear surface power, as well as advanced chemical propulsion can significantly enhance these scenarios. Robotic and human outer planet exploration options are described in many detailed and extensive studies. Nuclear propulsion options for fast trips to the outer planets are discussed. To refuel such vehicles, atmospheric mining in the outer solar system has also been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as helium 3 (3He) and hydrogen (H2) can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and H2 (deuterium, etc.) were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses have investigated resource capturing aspects of atmospheric mining in the outer solar system. These analyses included the gas capturing rate, storage options, and different methods of direct use of the captured gases. While capturing 3He, large amounts of hydrogen and 4He are produced. With these two additional gases, the potential for fueling small and large fleets of additional exploration and exploitation vehicles exists.

Performance potential of gas-core and fusion rockets - A mission applications survey.

NASA Technical Reports Server (NTRS)

Fishbach, L. H.; Willis, E. A., Jr.

1971-01-01

This paper reports an evaluation of the performance potential of five nuclear rocket engines for four mission classes. These engines are: the regeneratively cooled gas-core nuclear rocket; the light bulb gas-core nuclear rocket; the space-radiator cooled gas-core nuclear rocket; the fusion rocket; and an advanced solid-core nuclear rocket which is included for comparison. The missions considered are: earth-to-orbit launch; near-earth space missions; close interplanetary missions; and distant interplanetary missions. For each of these missions, the capabilities of each rocket engine type are compared in terms of payload ratio for the earth launch mission or by the initial vehicle mass in earth orbit for space missions (a measure of initial cost). Other factors which might determine the engine choice are discussed. It is shown that a 60 day manned round trip to Mars is conceivable.-

Implications of the Homogeneous Nucleation Barrier for Top-Down Crystallization in Mercury's Core

NASA Astrophysics Data System (ADS)

Huguet, L.; Hauck, S. A.; Van Orman, J. A.; Jing, Z.

2018-05-01

Crystallization of solids in planetary cores depends both on ambient temperatures falling below the liquidus and on the ability to nucleate crystal growth. We discuss the implications of the nucleation barrier for thermal evolution of Mercury's core.

Thermal barrier and support for nuclear reactor fuel core

DOEpatents

Betts, Jr., William S.; Pickering, J. Larry; Black, William E.

1987-01-01

A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

Heater Development, Fabrication, and Testing: Analysis of Fabricated Heaters

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Dickens, R. E.; Farmer, J. T.; Davis, J. D.; Adams, M. R.; Martin, J. J.; Webster, K. L.

2008-01-01

Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.

An integral nuclear power and propulsion system concept

NASA Astrophysics Data System (ADS)

Choong, Phillip T.; Teofilo, Vincent L.; Begg, Lester L.; Dunn, Charles; Otting, William

An integral space power concept provides both the electrical power and propulsion from a common heat source and offers superior performance capabilities over conventional orbital insertion using chemical propulsion systems. This paper describes a hybrid (bimodal) system concept based on a proven, inherently safe solid fuel form for the high temperature reactor core operation and rugged planar thermionic energy converter for long-life steady state electric power production combined with NERVA-based rocket technology for propulsion. The integral system is capable of long-life power operation and multiple propulsion operations. At an optimal thrust level, the integral system can maintain the minimal delta-V requirement while minimizing the orbital transfer time. A trade study comparing the overall benefits in placing large payloads to GEO with the nuclear electric propulsion option shows superiority of nuclear thermal propulsion. The resulting savings in orbital transfer time and the substantial reduction of overall lift requirement enables the use of low-cost launchers for several near-term military satellite missions.

Nuclear Engine System Simulation (NESS). Version 2.0: Program user's guide

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman

1993-01-01

This Program User's Guide discusses the Nuclear Thermal Propulsion (NTP) engine system design features and capabilities modeled in the Nuclear Engine System Simulation (NESS): Version 2.0 program (referred to as NESS throughout the remainder of this document), as well as its operation. NESS was upgraded to include many new modeling capabilities not available in the original version delivered to NASA LeRC in Dec. 1991, NESS's new features include the following: (1) an improved input format; (2) an advanced solid-core NERVA-type reactor system model (ENABLER 2); (3) a bleed-cycle engine system option; (4) an axial-turbopump design option; (5) an automated pump-out turbopump assembly sizing option; (6) an off-design gas generator engine cycle design option; (7) updated hydrogen properties; (8) an improved output format; and (9) personal computer operation capability. Sample design cases are presented in the user's guide that demonstrate many of the new features associated with this upgraded version of NESS, as well as design modeling features associated with the original version of NESS.

Some remarks on the early evolution of Enceladus

NASA Astrophysics Data System (ADS)

Czechowski, Leszek

2014-12-01

Thermal history of Enceladus is investigated from the beginning of accretion to formation of its core (~400 My). We consider model with solid state convection (in a solid layer) as well as liquid state convection (in molten parts of the satellite). The numerical model of convection uses full conservative finite difference method. The roles of two modes of convection are considered using the parameterized theory of convection. The following heat sources are included: short lived and long lived radioactive isotopes, accretion, serpentinization, and phase changes. Heat transfer processes are: conduction, solid state convection, and liquid state convection. It is found that core formation was completed only when liquid state convection had slowed down. Eventually, the porous core with pores filled with water was formed. Recent data concerning gravity field of Enceladus confirm low density of the core. We investigated also thermal history for different values of the following parameters: time of beginning of accretion tini, duration of accretion tacr, viscosity of ice close to the melting point ηm, activation energy in formula for viscosity E, thermal conductivity of silicate component ksil, ammonia content XNH3, and energy of serpentinization cserp. All these parameters are important for evolution, but not dramatic differences are found for realistic values. Moreover, the hypothesis of proto-Enceladus (stating that initially Enceladus was substantially larger) is considered and thermal history of such body is calculated. The last subject is the Mimas-Enceladus paradox. Comparison of thermal models of Mimas and Enceladus indicates that period favorable for 'excited path of evolution' was significantly shorter for Mimas than for Enceladus.

Symmetrical N-acylsubstituted dihydrazones containing bithiophene core - Photophysical, electrochemical and thermal characterization

NASA Astrophysics Data System (ADS)

Jarczyk-Jedryka, Anna; Filapek, Michal; Malecki, Grzegorz; Kula, Slawomir; Janeczek, Henryk; Boharewicz, Bartosz; Iwan, Agnieszka; Schab-Balcerzak, Ewa

2016-04-01

Four symmetrical N-acylsubstituted dihydrazones containing bithiophene core were synthesized from condensation of 2,2‧-bithiophene-5,5‧-dicarboxyaldehyde with benzoic, isonicotinoyl, 2-thiophenic and 2-furoic hydrazide. The obtained compounds were characterized through the data from 1H nuclear magnetic resonance spectroscopy (NMR), infrared spectroscopy (IR), elemental analysis, UV-vis absorption spectroscopy, photoluminescence (PL), cyclic voltammetry (CV) and differential pulse voltammetry (DPV) measurements. Additionally, the electronic properties including orbital energies and resulting energy gaps were calculated by density functional theory (DFT). Their thermal behavior was investigated by thermogravimetric analysis (TGA) and differential scanning calorimetry (DSC). They were thermal sable up to 320 °C. The prepared N-acylsubstituted dihydrazones emitted light with λem in the range of 499-530 nm in solution, whereas, in solid state as blend with PMMA blue emission was observed. They undergo quasi-reversible and irreversible electrochemical reduction and oxidation processes, respectively. Additionally, the selected compounds were tested preliminary as component of active layer in organic photovoltaic cells. The highest value of power conversion efficiency, equal to 1.68% under simulated 100 mW/cm2 AM 1.5G irradiation was found for device with the architecture ITO/PEDOT:PSS/P3HT:PCBM:FBTH (1:2:2)/Al.

Atmospheric Mining in the Outer Solar System: Resource Capturing, Storage, and Utilization

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2012-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and hydrogen can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and hydrogen (deuterium, etc.) were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate for hydrogen helium 4 and helium 3, storage options, and different methods of direct use of the captured gases. Additional supporting analyses were conducted to illuminate vehicle sizing and orbital transportation issues.

METHOD AND APPARATUS FOR EARTH PENETRATION

DOEpatents

Adams, W.M.

1963-12-24

A nuclear reactor apparatus for penetrating into the earth's crust is described. The apparatus comprises a cylindrical nuclear core operating at a temperature that is higher than the melting temperature of rock. A high-density ballast member is coupled to the nuclear core such that the overall density of the core-ballast assembly is greater than the density of molten rock. The nuclear core is thermally insulated so that its heat output is constrained to flow axially, with radial heat flow being minimized. In operation, the apparatus is placed in contact with the earth's crust at the point desired to be penetrated. The heat output of the reactor melts the underlying rock, and the apparatus sinks through the resulting magma. The fuel loading of the reactor core determines the ultimate depth of crust penetration. (AEC)

System Design for a Nuclear Electric Spacecraft Utilizing Out-of-core Thermionic Conversion

NASA Technical Reports Server (NTRS)

Estabrook, W. C.; Phillips, W. M.; Hsieh, T.

1976-01-01

Basic guidelines are presented for a nuclear space power system which utilizes heat pipes to transport thermal power from a fast nuclear reactor to an out of core thermionic converter array. Design parameters are discussed for the nuclear reactor, heat pipes, thermionic converters, shields (neutron and gamma), waste heat rejection systems, and the electrical bus bar-cable system required to transport the high current/low voltage power to the processing equipment. Dimensions are compatible with shuttle payload bay constraints.

Iron snow in the Martian core?

NASA Astrophysics Data System (ADS)

Davies, Christopher J.; Pommier, Anne

2018-01-01

The decline of Mars' global magnetic field some 3.8-4.1 billion years ago is thought to reflect the demise of the dynamo that operated in its liquid core. The dynamo was probably powered by planetary cooling and so its termination is intimately tied to the thermochemical evolution and present-day physical state of the Martian core. Bottom-up growth of a solid inner core, the crystallization regime for Earth's core, has been found to produce a long-lived dynamo leading to the suggestion that the Martian core remains entirely liquid to this day. Motivated by the experimentally-determined increase in the Fe-S liquidus temperature with decreasing pressure at Martian core conditions, we investigate whether Mars' core could crystallize from the top down. We focus on the "iron snow" regime, where newly-formed solid consists of pure Fe and is therefore heavier than the liquid. We derive global energy and entropy equations that describe the long-timescale thermal and magnetic history of the core from a general theory for two-phase, two-component liquid mixtures, assuming that the snow zone is in phase equilibrium and that all solid falls out of the layer and remelts at each timestep. Formation of snow zones occurs for a wide range of interior and thermal properties and depends critically on the initial sulfur concentration, ξ0. Release of gravitational energy and latent heat during growth of the snow zone do not generate sufficient entropy to restart the dynamo unless the snow zone occupies at least 400 km of the core. Snow zones can be 1.5-2 Gyrs old, though thermal stratification of the uppermost core, not included in our model, likely delays onset. Models that match the available magnetic and geodetic constraints have ξ0 ≈ 10% and snow zones that occupy approximately the top 100 km of the present-day Martian core.

Heat exchange studies on coconut oil cells as thermal energy storage for room thermal conditioning

NASA Astrophysics Data System (ADS)

Sutjahja, I. M.; Putri, Widya A.; Fahmi, Z.; Wonorahardjo, S.; Kurnia, D.

2017-07-01

As reported by many thermal environment experts, room air conditioning might be controlled by thermal mass system. In this paper we discuss the performance of coconut oil cells as room thermal energy storage. The heat exchange mechanism of coconut oil (CO) which is one of potential organic Phase Change Material (PCM) is studied based on the results of temperature measurements in the perimeter and core parts of cells. We found that the heat exchange performance, i.e. heat absorption and heat release processes of CO cells are dominated by heat conduction in the sensible solid from the higher temperature perimeter part to the lower temperature core part and heat convection during the solid-liquid phase transition and sensible liquid phase. The capability of heat absorption as measured by the reduction of air temperature is not influenced by CO cell size. Besides that, the application of CO as the thermal mass has to be accompanied by air circulation to get the cool sensation of the room’s occupants.

Nuclear core positioning system

DOEpatents

Garkisch, Hans D.; Yant, Howard W.; Patterson, John F.

1979-01-01

A structural support system for the core of a nuclear reactor which achieves relatively restricted clearances at operating conditions and yet allows sufficient clearance between fuel assemblies at refueling temperatures. Axially displaced spacer pads having variable between pad spacing and a temperature compensated radial restraint system are utilized to maintain clearances between the fuel elements. The core support plates are constructed of metals specially chosen such that differential thermal expansion produces positive restraint at operating temperatures.

The evolution of the moon and the terrestrial planets

NASA Technical Reports Server (NTRS)

Toksoez, M. N.; Johnston, D. H.

1974-01-01

The thermal evolutions of the Moon, Mars, Venus and Mercury are calculated theoretically starting from cosmochemical condensation models. An assortment of geological, geochemical and geophysical data are used to constrain both the present day temperatures and the thermal histories of the planets' interiors. Such data imply that the planets were heated during or shortly after formation and that all the terrestrial planets started their differentiations early in their history. The moon, smallest in size, is characterized as a differentiated body with a crust, a thick solid mantle and an interior region which may be partially molten. Mars, intermediate in size, is assumed to have differentiated an Fe-FeS core. Venus is characterized as a planet not unlike the earth in many respects. Core formation has occurred probably during the first billion years after the formation. Mercury, which probably has a large core, may have a 500 km thick solid lithosphere and a partially molten core if it is assumed that some heat sources exist in the core.

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Mathematical Modeling Of A Nuclear/Thermionic Power Source

NASA Technical Reports Server (NTRS)

Vandersande, Jan W.; Ewell, Richard C.

1992-01-01

Report discusses mathematical modeling to predict performance and lifetime of spacecraft power source that is integrated combination of nuclear-fission reactor and thermionic converters. Details of nuclear reaction, thermal conditions in core, and thermionic performance combined with model of swelling of fuel.

Measurement of thermal diffusivity of depleted uranium metal microspheres

NASA Astrophysics Data System (ADS)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-03-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Mirror fusion propulsion system: A performance comparison with alternate propulsion systems for the manned Mars Mission

NASA Technical Reports Server (NTRS)

Schulze, Norman R.; Carpenter, Scott A.; Deveny, Marc E.; Oconnell, T.

1993-01-01

The performance characteristics of several propulsion technologies applied to piloted Mars missions are compared. The characteristics that are compared are Initial Mass in Low Earth Orbit (IMLEO), mission flexibility, and flight times. The propulsion systems being compared are both demonstrated and envisioned: Chemical (or Cryogenic), Nuclear Thermal Rocket (NTR) solid core, NTR gas core, Nuclear Electric Propulsion (NEP), and a mirror fusion space propulsion system. The proposed magnetic mirror fusion reactor, known as the Mirror Fusion Propulsion System (MFPS), is described. The description is an overview of a design study that was conducted to convert a mirror reactor experiment at Lawrence Livermore National Lab (LLNL) into a viable space propulsion system. Design principles geared towards minimizing mass and maximizing power available for thrust are identified and applied to the LLNL reactor design, resulting in the MFPS. The MFPS' design evolution, reactor and fuel choices, and system configuration are described. Results of the performance comparison shows that the MFPS minimizes flight time to 60 to 90 days for flights to Mars while allowing continuous return-home capability while at Mars. Total MFPS IMLEO including propellant and payloads is kept to about 1,000 metric tons.

Mirror fusion propulsion system - A performance comparison with alternate propulsion systems for the manned Mars mission

NASA Technical Reports Server (NTRS)

Deveny, M.; Carpenter, S.; O'Connell, T.; Schulze, N.

1993-01-01

The performance characteristics of several propulsion technologies applied to piloted Mars missions are compared. The characteristics that are compared are Initial Mass in Low Earth Orbit (IMLEO), mission flexibility, and flight times. The propulsion systems being compared are both demonstrated and envisioned: Chemical (or Cryogenic), Nuclear Thermal Rocket (NTR) solid core, NTR gas core, Nuclear Electric Propulsion (NEP), and a mirror fusion space propulsion system. The proposed magnetic mirror fusion reactor, known as the Mirror Fusion Propulsion System (MFPS), is described. The description is an overview of a design study that was conducted to convert a mirror reactor experiment at Lawrence Livermore National Lab (LLNL) into a viable space propulsion system. Design principles geared towards minimizing mass and maximizing power available for thrust are identified and applied to the LLNL reactor design, resulting in the MFPS. The MFPS' design evolution, reactor and fuel choices, and system configuration are described. Results of the performance comparison shows that the MFPS minimizes flight time to 60 to 90 days for flights to Mars while allowing continuous return-home capability while at Mars. Total MFPS IMLEO including propellant and payloads is kept to about 1,000 metric tons.

Mechanical design

NASA Technical Reports Server (NTRS)

1976-01-01

Design concepts for a 1000 mw thermal stationary power plant employing the UF6 fueled gas core breeder reactor are examined. Three design combinations-gaseous UF6 core with a solid matrix blanket, gaseous UF6 core with a liquid blanket, and gaseous UF6 core with a circulating blanket were considered. Results show the gaseous UF6 core with a circulating blanket was best suited to the power plant concept.

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures inmore » the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.« less

Feasibility study of full-reactor gas core demonstration test

NASA Technical Reports Server (NTRS)

Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

1973-01-01

Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

Chemical Convention in the Lunar Core from Melting Experiments on the Ironsulfur System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, J.; Liu, J.; Chen, B.

2012-03-26

By reanalyzing Apollo lunar seismograms using array-processing methods, a recent study suggests that the Moon has a solid inner core and a fluid outer core, much like the Earth. The volume fraction of the lunar inner core is 38%, compared with 4% for the Earth. The pressure at the Moon's core-mantle boundary is 4.8 GPa, and that at the ICB is 5.2 GPa. The partially molten state of the lunar core provides constraints on the thermal and chemical states of the Moon: The temperature at the inner core boundary (ICB) corresponds to the liquidus of the outer core composition, andmore » the mass fraction of the solid core allows us to infer the bulk composition of the core from an estimated thermal profile. Moreover, knowledge on the extent of core solidification can be used to evaluate the role of chemical convection in the origin of early lunar core dynamo. Sulfur is considered an antifreeze component in the lunar core. Here we investigate the melting behavior of the Fe-S system at the pressure conditions of the lunar core, using the multi-anvil apparatus and synchrotron and laboratory-based analytical methods. Our goal is to understand compositionally driven convection in the lunar core and assess its role in generating an internal magnetic field in the early history of the Moon.« less

Dynamic nuclear polarization assisted spin diffusion for the solid effect case.

PubMed

Hovav, Yonatan; Feintuch, Akiva; Vega, Shimon

2011-02-21

The dynamic nuclear polarization (DNP) process in solids depends on the magnitudes of hyperfine interactions between unpaired electrons and their neighboring (core) nuclei, and on the dipole-dipole interactions between all nuclei in the sample. The polarization enhancement of the bulk nuclei has been typically described in terms of a hyperfine-assisted polarization of a core nucleus by microwave irradiation followed by a dipolar-assisted spin diffusion process in the core-bulk nuclear system. This work presents a theoretical approach for the study of this combined process using a density matrix formalism. In particular, solid effect DNP on a single electron coupled to a nuclear spin system is considered, taking into account the interactions between the spins as well as the main relaxation mechanisms introduced via the electron, nuclear, and cross-relaxation rates. The basic principles of the DNP-assisted spin diffusion mechanism, polarizing the bulk nuclei, are presented, and it is shown that the polarization of the core nuclei and the spin diffusion process should not be treated separately. To emphasize this observation the coherent mechanism driving the pure spin diffusion process is also discussed. In order to demonstrate the effects of the interactions and relaxation mechanisms on the enhancement of the nuclear polarization, model systems of up to ten spins are considered and polarization buildup curves are simulated. A linear chain of spins consisting of a single electron coupled to a core nucleus, which in turn is dipolar coupled to a chain of bulk nuclei, is considered. The interaction and relaxation parameters of this model system were chosen in a way to enable a critical analysis of the polarization enhancement of all nuclei, and are not far from the values of (13)C nuclei in frozen (glassy) organic solutions containing radicals, typically used in DNP at high fields. Results from the simulations are shown, demonstrating the complex dependences of the DNP-assisted spin diffusion process on variations of the relevant parameters. In particular, the effect of the spin lattice relaxation times on the polarization buildup times and the resulting end polarization are discussed, and the quenching of the polarizations by the hyperfine interaction is demonstrated.

Long Duration Hot Hydrogen Exposure of Nuclear Thermal Rocket Materials

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Hickman, Robert; Dobson, Chris; Clifton, Scooter

2007-01-01

An arc-heater driven hyper-thermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to .produce high-temperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low cost test facility for the purpose of investigating and characterizing candidate fuel/structural materials and improving associated processing/fabrication techniques. Design and engineering development efforts are fully summarized, and facility operating characteristics are reported as determined from a series of baseline performance mapping runs and long duration capability demonstration tests.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Realistic Testing of the Safe Affordable Fission Engine (SAFE-100) Thermal Simulator Using Fiber Bragg Gratings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stinson-Bagby, Kelly L.; Fielder, Robert S.; Van Dyke, Melissa K.

2004-02-04

The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. Distributed high temperature measurements were made with 20 FBG temperature sensors installed in the SAFE-100 thermal simulator at the NASA Marshal Space Flight Center. Experiments were performed at temperatures approaching 800 deg. C and 1150 deg. C for characterization studies of the SAFE-100 core. Temperature profiles were successfully generated for the core during temperature increases and decreases. Related tests in the SAFE-100 successfully provided strain measurement data.

Nuclear reactor safety device

DOEpatents

Hutter, E.

1983-08-15

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

The Rationale/Benefits of Nuclear Thermal Rocket Propulsion for NASA's Lunar Space Transportation System

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1994-01-01

The solid core nuclear thermal rocket (NTR) represents the next major evolutionary step in propulsion technology. With its attractive operating characteristics, which include high specific impulse (approximately 850-1000 s) and engine thrust-to-weight (approximately 4-20), the NTR can form the basis for an efficient lunar space transportation system (LTS) capable of supporting both piloted and cargo missions. Studies conducted at the NASA Lewis Research Center indicate that an NTR-based LTS could transport a fully-fueled, cargo-laden, lunar excursion vehicle to the Moon, and return it to low Earth orbit (LEO) after mission completion, for less initial mass in LEO than an aerobraked chemical system of the type studied by NASA during its '90-Day Study.' The all-propulsive NTR-powered LTS would also be 'fully reusable' and would have a 'return payload' mass fraction of approximately 23 percent--twice that of the 'partially reusable' aerobraked chemical system. Two NTR technology options are examined--one derived from the graphite-moderated reactor concept developed by NASA and the AEC under the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) programs, and a second concept, the Particle Bed Reactor (PBR). The paper also summarizes NASA's lunar outpost scenario, compares relative performance provided by different LTS concepts, and discusses important operational issues (e.g., reusability, engine 'end-of life' disposal, etc.) associated with using this important propulsion technology.

Thermal Evolution and Crystallisation Regimes of the Martian Core

NASA Astrophysics Data System (ADS)

Davies, C. J.; Pommier, A.

2015-12-01

Though it is accepted that Mars has a sulfur-rich metallic core, its chemical and physical state as well as its time-evolution are still unconstrained and debated. Several lines of evidence indicate that an internal magnetic field was once generated on Mars and that this field decayed around 3.7-4.0 Gyrs ago. The standard model assumes that this field was produced by a thermal (and perhaps chemical) dynamo operating in the Martian core. We use this information to construct parameterized models of the Martian dynamo in order to place constraints on the thermochemical evolution of the Martian core, with particular focus on its crystallization regime. Considered compositions are in the FeS system, with S content ranging from ~10 and 16 wt%. Core radius, density and CMB pressure are varied within the errors provided by recent internal structure models that satisfy the available geodetic constraints (planetary mass, moment of inertia and tidal Love number). We also vary the melting curve and adiabat, CMB heat flow and thermal conductivity. Successful models are those that match the dynamo cessation time and fall within the bounds on present-day CMB temperature. The resulting suite of over 500 models suggest three possible crystallization regimes: growth of a solid inner core starting at the center of the planet; freezing and precipitation of solid iron (Fe- snow) from the core-mantle boundary (CMB); and freezing that begins midway through the core. Our analysis focuses on the effects of core properties that are expected to be constrained during the forthcoming Insight mission.

Summary of the thermal evaluation of LWBR (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lerner, S.; McWilliams, K.D.; Stout, J.W.

1980-03-01

This report describes the thermal evaluation of the core for the Shippingport Light Water Breeder Reactor. This core contains unique thermal-hydraulic features such as (1) close rod-to-rod proximity, (2) an open-lattice array of fuel rods with two different diameters and rod-to-rod spacings in the same flow region, (3) triplate orifices located at both the entrance and exit of fuel modules and (4) a hydraulically-balanced movable-fuel system coupled with (5) axial-and-radial fuel zoning for reactivity control. Performance studies used reactor thermal principles such as the hot-and-nominal channel concept and related nuclear/engineering design allowances. These were applied to models of three-dimensional roddedmore » arrays comprising the core fuel regions.« less

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

NASA Technical Reports Server (NTRS)

Stanojev, B. J.; Houts, M.

2004-01-01

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Heat pipe nuclear reactor for space power

NASA Technical Reports Server (NTRS)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

NASA Technical Reports Server (NTRS)

Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

2013-01-01

In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

Nuclear propulsion - A vital technology for the exploration of Mars and the planets beyond

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1989-01-01

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the only other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulse of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class spaceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.

Nuclear propulsion: a vital technology for the exploration of Mars and the planets beyond

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borowski, S.K.

1988-01-01

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the olny other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-Earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulsemore » of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class speceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.« less

Nuclear propulsion: A vital technology for the exploration of Mars and the planets beyond

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1988-01-01

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the olny other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-Earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulse of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class speceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1993-10-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

Bimodal Nuclear Thermal Rocket Analysis Developments

NASA Technical Reports Server (NTRS)

Belair, Michael; Lavelle, Thomas; Saimento, Charles; Juhasz, Albert; Stewart, Mark

2014-01-01

Nuclear thermal propulsion has long been considered an enabling technology for human missions to Mars and beyond. One concept of operations for these missions utilizes the nuclear reactor to generate electrical power during coast phases, known as bimodal operation. This presentation focuses on the systems modeling and analysis efforts for a NERVA derived concept. The NERVA bimodal operation derives the thermal energy from the core tie tube elements. Recent analysis has shown potential temperature distributions in the tie tube elements that may limit the thermodynamic efficiency of the closed Brayton cycle used to generate electricity with the current design. The results of this analysis are discussed as well as the potential implications to a bimodal NERVA type reactor.

Arc-Heater Facility for Hot Hydrogen Exposure of Nuclear Thermal Rocket Materials

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Wang,Ten-See; Hickman, Robert; Panda, Binayak; Dobson, Chris; Osborne, Robin; Clifton, Scooter

2006-01-01

A hyper-thermal environment simulator is described for hot hydrogen exposure of nuclear thermal rocket material specimens and component development. This newly established testing capability uses a high-power, multi-gas, segmented arc-heater to produce high-temperature pressurized hydrogen flows representative of practical reactor core environments and is intended to serve. as a low cost test facility for the purpose of investigating and characterizing candidate fueUstructura1 materials and improving associated processing/fabrication techniques. Design and development efforts are thoroughly summarized, including thermal hydraulics analysis and simulation results, and facility operating characteristics are reported, as determined from a series of baseline performance mapping tests.

Thermal equation of state of hcp-iron: Constraint on the density deficit of Earth's solid inner core: THERMAL EQUATION OF STATE OF HCP-IRON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fei, Yingwei; Murphy, Caitlin; Shibazaki, Yuki

We conducted high-pressure experiments on hexagonal close packed iron (hcp-Fe) in MgO, NaCl, and Ne pressure-transmitting media and found general agreement among the experimental data at 300 K that yield the best fitted values of the bulk modulus K 0 = 172.7(±1.4) GPa and its pressure derivative K 0'= 4.79(±0.05) for hcp-Fe, using the third-order Birch-Murnaghan equation of state. Using the derived thermal pressures for hcp-Fe up to 100 GPa and 1800 K and previous shockwave Hugoniot data, we developed a thermal equation of state of hcp-Fe. The thermal equation of state of hcp-Fe is further used to calculate themore » densities of iron along adiabatic geotherms to define the density deficit of the inner core, which serves as the basis for developing quantitative composition models of the Earth's inner core. We determine the density deficit at the inner core boundary to be 3.6%, assuming an inner core boundary temperature of 6000 K.« less

Earth's inner core nucleation paradox

NASA Astrophysics Data System (ADS)

Huguet, Ludovic; Van Orman, James A.; Hauck, Steven A.; Willard, Matthew A.

2018-04-01

The conventional view of Earth's inner core is that it began to crystallize at Earth's center when the temperature dropped below the melting point of the iron alloy and has grown steadily since that time as the core continued to cool. However, this model neglects the energy barrier to the formation of the first stable crystal nucleus, which is commonly represented in terms of the critical supercooling required to overcome the barrier. Using constraints from experiments, simulations, and theory, we show that spontaneous crystallization in a homogeneous liquid iron alloy at Earth's core pressures requires a critical supercooling of order 1000 K, which is too large to be a plausible mechanism for the origin of Earth's inner core. We consider mechanisms that can lower the nucleation barrier substantially. Each has caveats, yet the inner core exists: this is the nucleation paradox. Heterogeneous nucleation on a solid metallic substrate tends to have a low energy barrier and offers the most straightforward solution to the paradox, but solid metal would probably have to be delivered from the mantle and such events are unlikely to have been common. A delay in nucleation, whether due to a substantial nucleation energy barrier, or late introduction of a low energy substrate, would lead to an initial phase of rapid inner core growth from a supercooled state. Such rapid growth may lead to distinctive crystallization texturing that might be observable seismically. It would also generate a spike in chemical and thermal buoyancy that could affect the geomagnetic field significantly. Solid metal introduced to Earth's center before it reached saturation could also provide a nucleation substrate, if large enough to escape complete dissolution. Inner core growth, in this case, could begin earlier and start more slowly than standard thermal models predict.

Solid state thermal rectifier

DOEpatents

None

2016-07-05

Thermal rectifiers using linear nanostructures as core thermal conductors have been fabricated. A high mass density material is added preferentially to one end of the nanostructures to produce an axially non-uniform mass distribution. The resulting nanoscale system conducts heat asymmetrically with greatest heat flow in the direction of decreasing mass density. Thermal rectification has been demonstrated for linear nanostructures that are electrical insulators, such as boron nitride nanotubes, and for nanostructures that are conductive, such as carbon nanotubes.

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Thermal control of high energy nuclear waste, space option. [mathematical models

NASA Technical Reports Server (NTRS)

Peoples, J. A.

1979-01-01

Problems related to the temperature and packaging of nuclear waste material for disposal in space are explored. An approach is suggested for solving both problems with emphasis on high energy density waste material. A passive cooling concept is presented which utilized conduction rods that penetrate the inner core. Data are presented to illustrate the effectiveness of the rods and the limit of their capability. A computerized thermal model is discussed and developed for the cooling concept.

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Coated ceramic breeder materials

DOEpatents

Tam, Shiu-Wing; Johnson, Carl E.

1987-01-01

A breeder material for use in a breeder blanket of a nuclear reactor is disclosed. The breeder material comprises a core material of lithium containing ceramic particles which has been coated with a neutron multiplier such as Be or BeO, which coating has a higher thermal conductivity than the core material.

Coated ceramic breeder materials

DOEpatents

Tam, Shiu-Wing; Johnson, Carl E.

1987-04-07

A breeder material for use in a breeder blanket of a nuclear reactor is disclosed. The breeder material comprises a core material of lithium containing ceramic particles which has been coated with a neutron multiplier such as Be or BeO, which coating has a higher thermal conductivity than the core material.

Processing of solid solution, mixed uranium/refractory metal carbides for advanced space nuclear power and propulsion systems

NASA Astrophysics Data System (ADS)

Knight, Travis Warren

Nuclear thermal propulsion (NTP) and space nuclear power are two enabling technologies for the manned exploration of space and the development of research outposts in space and on other planets such as Mars. Advanced carbide nuclear fuels have been proposed for application in space nuclear power and propulsion systems. This study examined the processing technologies and optimal parameters necessary to fabricate samples of single phase, solid solution, mixed uranium/refractory metal carbides. In particular, the pseudo-ternary carbide, UC-ZrC-NbC, system was examined with uranium metal mole fractions of 5% and 10% and corresponding uranium densities of 0.8 to 1.8 gU/cc. Efforts were directed to those methods that could produce simple geometry fuel elements or wafers such as those used to fabricate a Square Lattice Honeycomb (SLHC) fuel element and reactor core. Methods of cold uniaxial pressing, sintering by induction heating, and hot pressing by self-resistance heating were investigated. Solid solution, high density (low porosity) samples greater than 95% TD were processed by cold pressing at 150 MPa and sintering above 2600 K for times longer than 90 min. Some impurity oxide phases were noted in some samples attributed to residual gases in the furnace during processing. Also, some samples noted secondary phases of carbon and UC2 due to some hyperstoichiometric powder mixtures having carbon-to-metal ratios greater than one. In all, 33 mixed carbide samples were processed and analyzed with half bearing uranium as ternary carbides of UC-ZrC-NbC. Scanning electron microscopy, x-ray diffraction, and density measurements were used to characterize samples. Samples were processed from powders of the refractory mono-carbides and UC/UC 2 or from powders of uranium hydride (UH3), graphite, and refractory metal carbides to produce hypostoichiometric mixed carbides. Samples processed from the constituent carbide powders and sintered at temperatures above the melting point of UC showed signs of liquid phase sintering and were shown to be largely solid solutions. Pre-compaction of mixed carbide powders prior to sintering was shown to be necessary to achieve high densities. Hypostoichiometric, samples processed at 2500 K exhibited only the initial stage of sintering and solid solution formation. Based on these findings, a suggested processing methodology is proposed for producing high density, solid solution, mixed carbide fuels. Pseudo-binary, refractory carbide samples hot pressed at 3100 K and 6 MPa showed comparable densities (approximately 85% of the theoretical value) to samples processed by cold pressing and sintering at temperatures of 2800 K.

Fluid-mechanic/thermal interaction of a molten material and a decomposing solid

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larson, D.W.; Lee, D.O.

1976-12-01

Bench-scale experiments of a molten material in contact with a decomposing solid were conducted to gain insight into the expected interaction of a hot, molten reactor core with a concrete base. The results indicate that either of two regimes can occur: violent agitation and splattering of the melt or a very quiescent settling of the melt when placed in contact with the solid. The two regimes appear to be governed by the interface temperature condition. A conduction heat transfer model predicts the critical interface temperature with reasonable accuracy. In addition, a film thermal resistance model correlates well with the datamore » in predicting the time for a solid skin to form on the molten material.« less

United States Air Force Graduate Student Summer Support Program (1987). Program Technical Report. Volume 1.

DTIC Science & Technology

1987-12-01

developed for a large percentage of the participants in the Summer Faculty Research Program in 1979-1983 period through an AFOSR Minigrant Program . On 1...Analysis of a Bimodal Nuclear Rocket Core by Dav,, C. Carpenter ABSTRACT The framework for a general purpose finite element analysis code was developed ...to study the 2-D temperature distribution in a hot-channel S hexagonal fuel element in the core of a bimodal nuclear’ rocket. Prelim- inary thermal

Insights into Mercury's interior structure from geodesy measurements and global contraction

NASA Astrophysics Data System (ADS)

Rivoldini, A.; Van Hoolst, T.

2014-04-01

The measurements of the gravitational field of Mercury by MESSENGER [6] and improved measurements of the spin state of Mercury [3] provide important insights on its interior structure. In particular, these data give strong constraints on the radius and density of Mercury's core [5, 2]. However, present geodesy data do not provide strong constraints on the radius of the inner core. The data allow for models with a fully molten liquid core to models which have an inner core radius that is smaller than about 1760km [5], if it is assumed that sulfur is the only light element in the core. Models without an inner core are, however, at odds with the observed internally generated magnetic field of Mercury since Mercury's dynamo cannot operate by secular cooling alone at present. The present radius of the inner core depends mainly on Mercury's thermal state and light elements inside the core. Because of the secular cooling of the planet,the temperature inside the core drops below the liquidus temperature of the core material somewhere in the core and leads to the formation of an inner core and to the global contraction of the planet. The amount of contraction depends on the temperature decrease, on the thermal expansion of the materials inside the planet, and on the volume of crystallized liquid core alloy. In this study we use geodesy data, the recent estimate about the radial contraction of Mercury [1], and thermo-chemical evolution calculations in order to improve our knowledge about Mercury's inner core radius and thermal state. Since data from remote sensing of Mercury's surface [4] indicate that Mercury formed under reducing conditions we consider models that have sulfur and silicon as light elements inside their core. Unlike sulfur, which does almost not partition into solid iron under Mercury's core pressure and temperature conditions, silicon partitions virtually equally between solid and liquid iron. As a consequence, the density difference between the liquid and the crystallized material is smaller than for sulfur as only light element inside the core and therefore, for a given inner core radius the contraction of the planet is likely smaller.

Survey of Current and Next Generation Space Power Technologies

DTIC Science & Technology

2006-06-26

different thermodynamic cycles, such as the Brayton, Rankine, and Stirling cycles, alkali metal thermal electric converters ( AMTEC ) and thermionic...efficiencies @ 1700K. The primary issue with this system is the integration of the converter technology into the nuclear reactor core. AMTEC (static...Alkali metal thermal to electric converters ( AMTECs ) are thermally powered electrochemical concentration cells that convert heat energy directly to DC

First In-Core Simultaneous Measurements of Nuclear Heating and Thermal Neutron Flux Obtained With the Innovative Mobile Calorimeter CALMOS Inside the OSIRIS Reactor

NASA Astrophysics Data System (ADS)

Carcreff, Hubert; Salmon, Laurent; Bubendorff, Jacques; Lepeltier, Valérie

2016-10-01

Nuclear heating inside a MTR reactor has to be known in order to design and run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Calorimeter working modes, measurement procedures, main modeling and experimental results and expected advantages of this new technique have been already presented in previous papers. However, these first in-core measurements were not performed beyond 6 W · g-1, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 1014 n · cm-2 · s-1 and nuclear heating up to 12 W · g-1. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a comparison is made between results obtained by the probe calibration coefficient and the zero methods. Thermal neutron flux evaluation from SPND signal processing required a specific TRIPOLI-4 Monte Carlo calculation which has been performed with the precise CALMOS cell geometry. In addition, the Finite Element model for temperatures map prediction inside the calorimetric cell has been upgraded with recent experimental data obtained up to 12 W · g-1. Finally, the experience feedback led us to improvement perspectives. A second device is currently under manufacturing and main technical options are presented.

Constraining Mercury's interior structure with geodesy data and its present thermal state

NASA Astrophysics Data System (ADS)

Rivoldini, Attilio; Van Hoolst, Tim; Noack, Lena

2015-04-01

Recent measurements of Mercury's spin state and gravitational field supplemented by the assumption that the planet's core is made of iron and sulfur give strong constraints on its interior structure. In particular, they allow a precise determination of Mercury's core size and average mantle density. Present geodesy data do, however, almost not constrain the size of the inner core. Interior structure models with a fully molten liquid core as well as models with an inner core almost as large as the core agree with the observations. Additionally, the observed internally generated magnetic field of Mercury does not preclude the absence of an inner core, since remelting of iron snow inside the core could produce a sufficient buoyancy flux to drive magnetic field generation by compositional convection. Although sulfur is ubiquitously invoked as being the principal candidate light element in terrestrial planet's cores its abundance in the core depends on the redox conditions during planetary formation. Remote sensing data of Mercury's surface by MESSENGER indicate that Mercury formed under reducing conditions. As a consequence, substantial amounts of other light elements like for example silicon and carbon could be present together with sulfur inside Mercury's core. Compared to sulfur, which does almost not partition into solid iron at Mercury's core conditions, silicon partitions almost equally well between solid and liquid iron whereas a few percent of carbon can partition into solid iron. Therefore, compared to a pure iron-sulfur core, if silicon and carbon are present in the core the density jump at the inner-core outer-core boundary could be smaller and induce a large enough change in the inner-core flattening to alter Mercury's libration amplitude. Moreover, the presence of carbon together with sulfur further reduces the core solidus temperature, potentially delaying the onset of inner core formation. Finally, if both silicon and sulfur are present in sufficient quantities a thin layer much enriched in sulfur and depleted in silicon could form at the top of the core as a consequence of a large immiscibility region in liquid Fe-S-Si at Mercury's core conditions. The present radius of an inner core depends mainly on Mercury's thermal state and concentration of light elements inside the core. Because of the secular cooling of the planet, at a time in Mercury's evolution the temperature inside the core drops below the core liquidus temperature somewhere in the core, which can lead to the formation of an inner core and to the global contraction of the planet. The amount of contraction depends mainly on the temperature decrease, on the thermal expansion of the materials inside the planet, on the volume of crystallized iron-rich core liquid, and on the volume of crystallized crust. In this study we use geodesy data (88 day libration amplitude, polar moment of inertia, and tidal Love number), the recent estimate about the radial contraction of Mercury, and thermo-chemical evolution calculations taking into account the formation of the crust, a growing inner core, and modeling the formation of iron-rich snow in the core in order to improve our knowledge about Mercury's inner core radius and thermal state. Since data from remote sensing of Mercury's surface indicate that Mercury formed under reducing conditions we consider models that have sulfur, silicon, and carbon as light elements inside their core.

SAVY 4000 Container Filter Design Life and Extension Implementation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moore, Murray E.; Reeves, Kirk Patrick; Veirs, Douglas Kirk

The SAVY 4000 is a general purpose, reusable container for the storage of solid nuclear material inside a nuclear facility. The canister has a permitted loading for material with a thermal output not to exceed 25 watts. This wattage limit applies to all containers, regardless of their size.

Rocket Motor Joint Construction Including Thermal Barrier

NASA Technical Reports Server (NTRS)

Steinetz, Bruce M. (Inventor); Dunlap, Patrick H., Jr. (Inventor)

2002-01-01

A thermal barrier for extremely high temperature applications consists of a carbon fiber core and one or more layers of braided carbon fibers surrounding the core. The thermal barrier is preferably a large diameter ring, having a relatively small cross-section. The thermal barrier is particularly suited for use as part of a joint structure in solid rocket motor casings to protect low temperature elements such as the primary and secondary elastomeric O-ring seals therein from high temperature gases of the rocket motor. The thermal barrier exhibits adequate porosity to allow pressure to reach the radially outward disposed O-ring seals allowing them to seat and perform the primary sealing function. The thermal barrier is disposed in a cavity or groove in the casing joint, between the hot propulsion gases interior of the rocket motor and primary and secondary O-ring seals. The characteristics of the thermal barrier may be enhanced in different applications by the inclusion of certain compounds in the casing joint, by the inclusion of RTV sealant or similar materials at the site of the thermal barrier, and/or by the incorporation of a metal core or plurality of metal braids within the carbon braid in the thermal barrier structure.

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

NASA Technical Reports Server (NTRS)

Panda, Binayak; Hickman, Robert R.; Shah, Sandeep

2005-01-01

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Synthesis of composite nanoparticles using co-precipitation of a magnetic iron-oxide shell onto core nanoparticles

NASA Astrophysics Data System (ADS)

Primc, Darinka; Belec, Blaž; Makovec, Darko

2016-03-01

Composite nanoparticles can be synthesized by coating a shell made of one material onto core nanoparticles made of another material. Here we report on a novel method for coating a magnetic iron oxide onto the surface of core nanoparticles in an aqueous suspension. The method is based on the heterogeneous nucleation of an initial product of Fe3+/Fe2+ co-precipitation on the core nanoparticles. The close control of the supersaturation of the precipitating species required for an exclusively heterogeneous nucleation and the growth of the shell were achieved by immobilizing the reactive Fe3+ ions in a nitrate complex with urea ([Fe((CO(NH2)2)6](NO3)3) and by using solid Mg(OH)2 as the precipitating reagent. The slow thermal decomposition of the complex at 60 °C homogeneously releases the reactive Fe3+ ions into the suspension of the core nanoparticles. The key stage of the process is the thermal hydrolysis of the released Fe3+ ions prior to the addition of Mg(OH)2. The thermal hydrolysis results in the formation of γ-FeOOH, exclusively at the surfaces of the core nanoparticles. After the addition of the solid hydroxide Mg(OH)2, the pH increases and at pH 5.7 the Fe2+ precipitates and reacts with the γ-FeOOH to form magnetic iron oxide with a spinel structure (spinel ferrite) at the surfaces of the core nanoparticles. The proposed low-temperature method for the synthesis of composite nanoparticles is capable of forming well-defined interfaces between the two components, important for the coupling of the different properties. The procedure is environmentally friendly, inexpensive, and appropriate for scaling up to mass production.

GAS COOLED NUCLEAR REACTORS

DOEpatents

Long, E.; Rodwell, W.

1958-06-10

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

Induction simulation of gas core nuclear engine

NASA Technical Reports Server (NTRS)

Poole, J. W.; Vogel, C. E.

1973-01-01

The design, construction and operation of an induction heated plasma device known as a combined principles simulator is discussed. This device incorporates the major design features of the gas core nuclear rocket engine such as solid feed, propellant seeding, propellant injection through the walls, and a transpiration cooled, choked flow nozzle. Both argon and nitrogen were used as propellant simulating material, and sodium was used for fuel simulating material. In addition, a number of experiments were conducted utilizing depleted uranium as the fuel. The test program revealed that satisfactory operation of this device can be accomplished over a range of operating conditions and provided additional data to confirm the validity of the gas core concept.

Liquid Film Migration in Warm Formed Aluminum Brazing Sheet

NASA Astrophysics Data System (ADS)

Benoit, M. J.; Whitney, M. A.; Wells, M. A.; Jin, H.; Winkler, S.

2017-10-01

Warm forming has previously proven to be a promising manufacturing route to improve formability of Al brazing sheets used in automotive heat exchanger production; however, the impact of warm forming on subsequent brazing has not previously been studied. In particular, the interaction between liquid clad and solid core alloys during brazing through the process of liquid film migration (LFM) requires further understanding. Al brazing sheet comprised of an AA3003 core and AA4045 clad alloy, supplied in O and H24 tempers, was stretched between 0 and 12 pct strain, at room temperature and 523K (250 °C), to simulate warm forming. Brazeability was predicted through thermal and microstructure analysis. The rate of solid-liquid interactions was quantified using thermal analysis, while microstructure analysis was used to investigate the opposing processes of LFM and core alloy recrystallization during brazing. In general, liquid clad was consumed relatively rapidly and LFM occurred in forming conditions where the core alloy did not recrystallize during brazing. The results showed that warm forming could potentially impair brazeability of O temper sheet by extending the regime over which LFM occurs during brazing. No change in microstructure or thermal data was found for H24 sheet when the forming temperature was increased, and thus warm forming was not predicted to adversely affect the brazing performance of H24 sheet.

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« less

Alfven seismic vibrations of crustal solid-state plasma in quaking paramagnetic neutron star

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bastrukov, S.; Xu, R.-X.; Molodtsova, I.

2010-11-15

Magneto-solid-mechanical model of two-component, core-crust, paramagnetic neutron star responding to quake-induced perturbation by differentially rotational, torsional, oscillations of crustal electron-nuclear solid-state plasma about axis of magnetic field frozen in the immobile paramagnetic core is developed. Particular attention is given to the node-free torsional crust-against-core vibrations under combined action of Lorentz magnetic and Hooke's elastic forces; the damping is attributed to Newtonian force of shear viscose stresses in crustal solid-state plasma. The spectral formulas for the frequency and lifetime of this toroidal mode are derived in analytic form and discussed in the context of quasiperiodic oscillations of the x-ray outburst fluxmore » from quaking magnetars. The application of obtained theoretical spectra to modal analysis of available data on frequencies of oscillating outburst emission suggests that detected variability is the manifestation of crustal Alfven's seismic vibrations restored by Lorentz force of magnetic field stresses.« less

Path-integral simulation of solids.

PubMed

Herrero, C P; Ramírez, R

2014-06-11

The path-integral formulation of the statistical mechanics of quantum many-body systems is described, with the purpose of introducing practical techniques for the simulation of solids. Monte Carlo and molecular dynamics methods for distinguishable quantum particles are presented, with particular attention to the isothermal-isobaric ensemble. Applications of these computational techniques to different types of solids are reviewed, including noble-gas solids (helium and heavier elements), group-IV materials (diamond and elemental semiconductors), and molecular solids (with emphasis on hydrogen and ice). Structural, vibrational, and thermodynamic properties of these materials are discussed. Applications also include point defects in solids (structure and diffusion), as well as nuclear quantum effects in solid surfaces and adsorbates. Different phenomena are discussed, as solid-to-solid and orientational phase transitions, rates of quantum processes, classical-to-quantum crossover, and various finite-temperature anharmonic effects (thermal expansion, isotopic effects, electron-phonon interactions). Nuclear quantum effects are most remarkable in the presence of light atoms, so that especial emphasis is laid on solids containing hydrogen as a constituent element or as an impurity.

SPOC Benchmark Case: SNRE Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vishal Patel; Michael Eades; Claude Russel Joyner II

The Small Nuclear Rocket Engine (SNRE) was modeled in the Center for Space Nuclear Research’s (CSNR) Space Propulsion Optimization Code (SPOC). SPOC aims to create nuclear thermal propulsion (NTP) geometries quickly to perform parametric studies on design spaces of historic and new NTP designs. The SNRE geometry was modeled in SPOC and a critical core with a reasonable amount of criticality margin was found. The fuel, tie-tubes, reflector, and control drum masses were predicted rather well. These are all very important for neutronics calculations so the active reactor geometries created with SPOC can continue to be trusted. Thermal calculations ofmore » the average and hot fuel channels agreed very well. The specific impulse calculations used historically and in SPOC disagree so mass flow rates and impulses differed. Modeling peripheral and power balance components that do not affect nuclear characteristics of the core is not a feature of SPOC and as such, these components should continue to be designed using other tools. A full paper detailing the available SNRE data and comparisons with SPOC outputs will be submitted as a follow-up to this abstract.« less

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Late-time Cooling of Neutron Star Transients and the Physics of the Inner Crust

NASA Astrophysics Data System (ADS)

Deibel, Alex; Cumming, Andrew; Brown, Edward F.; Reddy, Sanjay

2017-04-01

An accretion outburst onto a neutron star transient heats the neutron star’s crust out of thermal equilibrium with the core. After the outburst, the crust thermally relaxes toward equilibrium with the neutron star core, and the surface thermal emission powers the quiescent X-ray light curve. Crust cooling models predict that thermal equilibrium of the crust will be established ≈ 1000 {days} into quiescence. Recent observations of the cooling neutron star transient MXB 1659-29, however, suggest that the crust did not reach thermal equilibrium with the core on the predicted timescale and continued to cool after ≈ 2500 {days} into quiescence. Because the quiescent light curve reveals successively deeper layers of the crust, the observed late-time cooling of MXB 1659-29 depends on the thermal transport in the inner crust. In particular, the observed late-time cooling is consistent with a low thermal conductivity layer near the depth predicted for nuclear pasta that maintains a temperature gradient between the neutron star’s inner crust and core for thousands of days into quiescence. As a result, the temperature near the crust-core boundary remains above the critical temperature for neutron superfluidity, and a layer of normal neutrons forms in the inner crust. We find that the late-time cooling of MXB 1659-29 is consistent with heat release from a normal neutron layer near the crust-core boundary with a long thermal time. We also investigate the effect of inner crust physics on the predicted cooling curves of the accreting transient KS 1731-260 and the magnetar SGR 1627-41.

Changes in porosity and organic matter phase distribution monitored by NMR relaxometry following hydrous pyrolysis under uniaxial confinement

USGS Publications Warehouse

Washburn, Kathryn E.; Birdwell, Justin E.; Lewan, Michael D.; Miller, Michael; Baez, Luis; Beeney, Ken; Sonnenberg, Steve

2013-01-01

Artificial maturation methods are used to induce changes in source rock thermal maturity without the uncertainties that arise when comparing natural samples from a particular basin that often represent different levels of maturation and different lithofacies. A novel uniaxial confinement clamp was used on Woodford Shale cores in hydrous pyrolysis experiments to limit sample expansion by simulating the effect of overburden present during thermal maturation in natural systems. These samples were then subjected to X-ray computed tomography (X-CT) imaging and low-field nuclear magnetic resonance (LF-NMR) relaxometry measurements. LF-NMR relaxometry is a noninvasive technique commonly used to measure porosity and pore-size distributions in fluid-filled porous media, but may also measure hydrogen present in hydrogen-bearing organic solids. Standard T1 and T2 relaxation distributions were determined and two dimensional T1-T2 correlation measurements were performed on the Woodford Shale cores. The T1-T2 correlations facilitate resolution of organic phases in the system. The changes observed in NMR-relaxation times correspond to bitumen and lighter hydrocarbon production that occur as source rock organic matter matures. The LF-NMR porosities of the core samples at maximum oil generation are significantly higher than porosities measured by other methods. This discrepancy likely arises from the measurement of highly viscous organic constituents in addition to fluid-filled porosity. An unconfined sample showed shorter relaxation times and lower porosity. This difference is attributed to the lack of fractures observed in the unconfined sample by X-CT.

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Thermal annihilation of photo-induced radicals following dynamic nuclear polarization to produce transportable frozen hyperpolarized 13C-substrates

PubMed Central

Capozzi, Andrea; Cheng, Tian; Boero, Giovanni; Roussel, Christophe; Comment, Arnaud

2017-01-01

Hyperpolarization via dynamic nuclear polarization (DNP) is pivotal for boosting magnetic resonance imaging (MRI) sensitivity and dissolution DNP can be used to perform in vivo real-time 13C MRI. The type of applications is however limited by the relatively fast decay time of the hyperpolarized spin state together with the constraint of having to polarize the 13C spins in a dedicated apparatus nearby but separated from the MRI magnet. We herein demonstrate that by polarizing 13C with photo-induced radicals, which can be subsequently annihilated using a thermalization process that maintains the sample temperature below its melting point, hyperpolarized 13C-substrates can be extracted from the DNP apparatus in the solid form, while maintaining the enhanced 13C polarization. The melting procedure necessary to transform the frozen solid into an injectable solution containing the hyperpolarized 13C-substrates can therefore be performed ex situ, up to several hours after extraction and storage of the polarized solid. PMID:28569840

Nuclear-Rocket Propulsion

NASA Technical Reports Server (NTRS)

Rom, Frank E.

1968-01-01

The three basic types of nuclear power-plants (solid, liquid, and gas core) are compared on the bases of performance potential and the status of current technology. The solid-core systems are expected to have impulses in the range of 850 seconds, any thrust level (as long as it is greater than 10,000 pounds (44,480 newtons)), and thrust-to-engine-weight ratios of 2 to 20 pounds per pound (19.7 to 197 newtons per kilogram). There is negligible or no fuel loss from the solid-core system. The solid-core system, of course, has had the most work done on it. Large-scale tests have been performed on a breadboard engine that has produced specific impulses greater than 700 seconds at thrust levels of about 50,000 pounds (222,000 newtons). The liquid-core reactor would be interesting in the specific impulse range of 1200 to 1500 seconds. Again, any thrust level can be obtained depending on how big or small the reactor is made. The thrust-to-engine weight ratio for these systems would be in the range of 1 to 10. The discouraging feature of the liquid-core system is the high fuel-loss ratio anticipated. Values of 0.01 to 0.1 pound (0.00454 to 0.0454 kilograms) or uranium loss per pound (0.454 kilograms) of hydrogen are expected, if impulses in the range of 1200 to 1500 seconds are desired. The gas-core reactor shows specific impulses in the range of 1500 to 2500 seconds. The thrust levels should be at least as high as the weight so that the thrust-to-weight ratio does not go below 1. Because the engine weight is not expected to be under 100,000 pounds (444,800 newtons), thrust levels higher than 100,000 pounds (448,000 newtons) are of interest. The thrust-to-engine weights, in that case, would run from 1 to 20 pounds per pound (9.8 to 19.7 kilograms). Gas-core reactors tend to be very large, and can have high thrust-to-weight ratios. As in the case of the liquid-core system, the fuel loss that will be attendant with gas cores as envisioned today will be rather high. The loss rates will be 0.01 to 0.1 pound of uranium (0.00454 to 0.0454 kilograms) for each pound (0.454 kilograms) of hydrogen.

Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

NASA Astrophysics Data System (ADS)

Khatry, Jivan

Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

Phase Equilibrium Experiments on Potential Lunar Core Compositions: Extension of Current Knowledge to Multi-Component (Fe-Ni-Si-S-C) Systems

NASA Technical Reports Server (NTRS)

Righter, K.; Pando, K.; Danielson, L.

2014-01-01

Numerous geophysical and geochemical studies have suggested the existence of a small metallic lunar core, but the composition of that core is not known. Knowledge of the composition can have a large impact on the thermal evolution of the core, its possible early dynamo creation, and its overall size and fraction of solid and liquid. Thermal models predict that the current temperature at the core-mantle boundary of the Moon is near 1650 K. Re-evaluation of Apollo seismic data has highlighted the need for new data in a broader range of bulk core compositions in the PT range of the lunar core. Geochemical measurements have suggested a more volatile-rich Moon than previously thought. And GRAIL mission data may allow much better constraints on the physical nature of the lunar core. All of these factors have led us to determine new phase equilibria experimental studies in the Fe-Ni-S-C-Si system in the relevant PT range of the lunar core that will help constrain the composition of Moon's core.

Mercury's core evolution

NASA Astrophysics Data System (ADS)

Deproost, Marie-Hélène; Rivoldini, Attilio; Van Hoolst, Tim

2016-10-01

Remote sensing data of Mercury's surface by MESSENGER indicate that Mercury formed under reducing conditions. As a consequence, silicon is likely the main light element in the core together with a possible small fraction of sulfur. Compared to sulfur, which does almost not partition into solid iron at Mercury's core conditions and strongly decreases the melting temperature, silicon partitions almost equally well between solid and liquid iron and is not very effective at reducing the melting temperature of iron. Silicon as the major light element constituent instead of sulfur therefore implies a significantly higher core liquidus temperature and a decrease in the vigor of compositional convection generated by the release of light elements upon inner core formation.Due to the immiscibility in liquid Fe-Si-S at low pressure (below 15 GPa), the core might also not be homogeneous and consist of an inner S-poor Fe-Si core below a thinner Si-poor Fe-S layer. Here, we study the consequences of a silicon-rich core and the effect of the blanketing Fe-S layer on the thermal evolution of Mercury's core and on the generation of a magnetic field.

Physics and potentials of fissioning plasmas for space power and propulsion

NASA Technical Reports Server (NTRS)

Thom, K.; Schwenk, F. C.; Schneider, R. T.

1976-01-01

Fissioning uranium plasmas are the nuclear fuel in conceptual high-temperature gaseous-core reactors for advanced rocket propulsion in space. A gaseous-core nuclear rocket would be a thermal reactor in which an enriched uranium plasma at about 10,000 K is confined in a reflector-moderator cavity where it is nuclear critical and transfers its fission power to a confining propellant flow for the production of thrust at a specific impulse up to 5000 sec. With a thrust-to-engine weight ratio approaching unity, the gaseous-core nuclear rocket could provide for propulsion capabilities needed for manned missions to the nearby planets and for economical cislunar ferry services. Fueled with enriched uranium hexafluoride and operated at temperatures lower than needed for propulsion, the gaseous-core reactor scheme also offers significant benefits in applications for space and terrestrial power. They include high-efficiency power generation at low specific mass, the burnup of certain fission products and actinides, the breeding of U-233 from thorium with short doubling times, and improved convenience of fuel handling and processing in the gaseous phase.

Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, Steven C.; Madey,Theodore E.; Haustein, Peter E.

2000-06-01

The purpose of this project is to deliver pertinent information that can be used to make rational decisions about the safety and treatment issues associated with dry storage of spent nuclear fuel materials. In particular, we will establish an understanding of: (1) water interactions with failed-fuel rods and metal-oxide materials; (2) the role of thermal processes and radiolysis (solid-state and interfacial) in the generation of potentially explosive mixtures of gaseous H2 and O2; and (3) the potential role of radiation-assisted corrosion during fuel rod storage.

Monterey Peninsula Water Supply Project. Supplemental Draft Environmental Impact Report/Statement II. Volume 2.

DTIC Science & Technology

1993-02-01

Fueled 7,634,0(X) 51 Geothermal 1,302,M(K) 9 Nuclear 2,160,(MX) 14 Total Thermal 11,096,(kM) 74 Hydroelectric 3,877,M(X) 26 Solar 0 0t Total Company...Nuclear 16,273,963 17 "Total Thermal 48,094,316 50 Hydroelectric 8,007,631 8 Solar 35 0 Total Company Generation 56,101,982 58 Helms Pumpback Energy...returnable beverage containers, prohibition of disposable diapers , and other measures to reduce the volume of the urban solid waste streams. Appeaidix 19-B

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Edwards, Michael J.

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

Multiple Experimental Efforts to Understand the Structure and Dynamics of Earth's Core

NASA Astrophysics Data System (ADS)

Fei, Y.; Han, L.; Bennett, N.; Hou, M.; Kuwayama, Y.; Huang, H.

2014-12-01

It requires integration of data from different types of high-pressure experiments to understand the structure and dynamics of Earth's core. In particular, measurements of physical properties and element partitioning in systems relevant to the core provide complementary data to narrow down the range of possible core compositions. We have performed both static and dynamic compression experiments and combined results from these with literature data to establish a reliable thermal equation of state of iron. This allows us to precisely determine the density deficit in the solid inner core. The combination of density and sound velocity measurements for both solid and liquid iron and its alloys provide tight constraints on the density deficit in the liquid outer core and the amount of sulphur required to match the geophysical observations. We then conducted element-partitioning experiments between solid and liquid iron in both multi-anvil apparatus and the laser-heated diamond-anvil cell to determine the sulphur, silicon, and oxygen partitioning between the liquid outer core and solid inner core. We present newly developed high-pressure experimental and nano-scale analytical techniques that allow us to simulate the conditions of the inner core boundary (ICB) and analyze the chemical compositions of coexisting phases in the recovered samples. We have established protocols to obtain high-quality partitioning data in the laser-heating diamond-anvil cell combined with FIB/SEM crossbeam technology. The partitioning data obtained up to at least 200 GPa provide additional criteria to explain the observed density and velocity jumps at the ICB.

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

NASA Astrophysics Data System (ADS)

Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

2001-02-01

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented. .

Intensity of geomagnetic field in the Precambrian and evolution of the Earth's deep interior

NASA Astrophysics Data System (ADS)

Smirnov, A. V.

2017-09-01

Reliable data on the paleointensity of the geomagnetic field can become an important source of information both about the mechanisms of generation of the field at present and in the past, and about the internal structure of the Earth, especially the structure and evolution of its core. Unfortunately, the reliability of these data remains a serious problem of paleomagnetic research because of the limitations of experimental methods, and the complexity and diversity of rocks and their magnetic carriers. This is true even for relatively "young" Phanerozoic rocks, but investigation of Precambrian rocks is associated with many additional difficulties. As a consequence, our current knowledge of paleointensity, especially in the Precambrian period, is still very limited. The data limitations do not preclude attempts to use the currently available paleointensity results to analyze the evolution and characteristics of the Earth's internal structure, such as the age of the Earth's solid inner core or thermal conductivity in the liquid core. However, such attempts require considerable caution in handling data. In particular, it has now been reliably established that some results on the Precambrian paleointensity overestimate the true paleofield strength. When the paleointensity overestimates are excluded from consideration, the range of the field strength changes in the Precambrian does not exceed the range of its variation in the Phanerozoic. This result calls into question recent assertions that the Earth's inner core formed in the Mesoproterozoic, about 1.3 billion years ago, triggering a statistically significant increase in the long-term average field strength. Instead, our analysis has shown that the quantity and quality of the currently available data on the Precambrian paleointensity are insufficient to estimate the age of the solid inner core and, therefore, cannot be useful for solving the problem of the thermal conductivity of the Earth's core. The data are consistent with very young or very "old" inner core ages and, correspondingly, with high or low values of core thermal conductivity.

Solid-State Division progress report for period ending March 31, 1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

Green, P.H.; Watson, D.M.

1983-09-01

Progress and activities are reported on: theoretical solid-state physics (surfaces; electronic, vibrational, and magnetic properties; particle-solid interactions; laser annealing), surface and near-surface properties of solids (surface, plasma-material interactions, ion implantation and ion-beam mixing, pulsed-laser and thermal processing), defects in solids (radiation effects, fracture, impurities and defects, semiconductor physics and photovoltaic conversion), transport properties of solids (fast-ion conductors, superconductivity, mass and charge transport in materials), neutron scattering (small-angle scattering, lattice dynamics, magnetic properties, structure and instrumentation), and preparation and characterization of research materials (growth and preparative methods, nuclear waste forms, special materials). (DLC)

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-22

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Chemical and nanometer-scale structure of kerogen and its change during thermal maturation investigated by advanced solid-state 13C NMR spectroscopy

USGS Publications Warehouse

Mao, J.; Fang, X.; Lan, Y.; Schimmelmann, A.; Mastalerz, Maria; Xu, L.; Schmidt-Rohr, K.

2010-01-01

We have used advanced and quantitative solid-state nuclear magnetic resonance (NMR) techniques to investigate structural changes in a series of type II kerogen samples from the New Albany Shale across a range of maturity (vitrinite reflectance R0 from 0.29% to 1.27%). Specific functional groups such as CH3, CH2, alkyl CH, aromatic CH, aromatic C-O, and other nonprotonated aromatics, as well as "oil prone" and "gas prone" carbons, have been quantified by 13C NMR; atomic H/C and O/C ratios calculated from the NMR data agree with elemental analysis. Relationships between NMR structural parameters and vitrinite reflectance, a proxy for thermal maturity, were evaluated. The aromatic cluster size is probed in terms of the fraction of aromatic carbons that are protonated (???30%) and the average distance of aromatic C from the nearest protons in long-range H-C dephasing, both of which do not increase much with maturation, in spite of a great increase in aromaticity. The aromatic clusters in the most mature sample consist of ???30 carbons, and of ???20 carbons in the least mature samples. Proof of many links between alkyl chains and aromatic rings is provided by short-range and long-range 1H-13C correlation NMR. The alkyl segments provide most H in the samples; even at a carbon aromaticity of 83%, the fraction of aromatic H is only 38%. While aromaticity increases with thermal maturity, most other NMR structural parameters, including the aromatic C-O fractions, decrease. Aromaticity is confirmed as an excellent NMR structural parameter for assessing thermal maturity. In this series of samples, thermal maturation mostly increases aromaticity by reducing the length of the alkyl chains attached to the aromatic cores, not by pronounced growth of the size of the fused aromatic ring clusters. ?? 2010 Elsevier Ltd. All rights reserved.

TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1996-12-31

The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient.

Measurements of thermophysical properties of solids at the Institute VINČA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milošević, Nenad, E-mail: nenadm@vinca.rs; Stepanić, Nenad, E-mail: nenad.s@vinca.rs; Terzić, Marijana, E-mail: marijanab@vinca.rs

2016-07-07

This paper presents the Metrological Laboratory for Thermophysical Quantities (MLTV) and its actual measurement possibilities. The MLTV is located in the Department of Thermal Engineering and Energy of the Institute of Nuclear Sciences VINČA in Serbia. It was founded in 1963, accredited by the National Accreditation Body in 2007 and became the national designated laboratory for thermophysical quantities and received the status of a EURAMET Associate Member in 2015. Today, the laboratory develops, maintains and disseminates traceability of different national standards, such as those for thermal conductivity of insulations and poorly conductive solid materials from 250 K to 350 K,more » thermal diffusivity of a large variety of solid materials from 200 K to 1450 K and specific heat and specific electrical resistivity from 250 K to 2400 K of electroconductive solid materials. Total hemispherical and spectral normal emissivity from 1200 K to 2400 K of electroconductive solid materials are also measured in the MLTV. The methods and experimental setups for the realization and measurement of all of these standards and quantities are described with corresponding examples.« less

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

SoLid Detector Technology

NASA Astrophysics Data System (ADS)

Labare, Mathieu

2017-09-01

SoLid is a reactor anti-neutrino experiment where a novel detector is deployed at a minimum distance of 5.5 m from a nuclear reactor core. The purpose of the experiment is three-fold: to search for neutrino oscillations at a very short baseline; to measure the pure 235U neutrino energy spectrum; and to demonstrate the feasibility of neutrino detectors for reactor monitoring. This report presents the unique features of the SoLid detector technology. The technology has been optimised for a high background environment resulting from low overburden and the vicinity of a nuclear reactor. The versatility of the detector technology is demonstrated with a 288 kg detector prototype which was deployed at the BR2 nuclear reactor in 2015. The data presented includes both reactor on, reactor off and calibration measurements. The measurement results are compared with Monte Carlo simulations. The 1.6t SoLid detector is currently under construction, with an optimised design and upgraded material technology to enhance the detector capabilities. Its deployement on site is planned for the begin of 2017 and offers the prospect to resolve the reactor anomaly within about two years.

First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert

2015-07-01

Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heatingmore » rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by different methods, the probe calibration coefficient and the zero method. Thermal neutron flux evaluation from the SPND signal processing required a specific TRIPOLI-4 Monte Carlo calculation which has been performed with the precise CALMOS cell geometry. In addition, the Finite Element model for temperatures map prediction inside the calorimetric cell has been upgraded with the recent experimental data obtained up to 12 W.g{sup -1}. The Kc coefficient, taking into account nonlinearities with regard to the calibration, has been reevaluated so as to make relevant measurements up to the nominal reactor power. Finally, the experience feedback acquired until now with this first CALMOS version led us to improvement perspectives. A second device is currently under manufacturing and main technical options chosen for this second version are presented. (authors)« less

Spin accumulation in thin Cs salts on contact with optically polarized Cs vapor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ishikawa, Kiyoshi

2011-09-15

The spin angular momentum accumulates in the Cs nuclei of salt on contact with optically pumped Cs vapor. The spin polarization in stable chloride as well as dissociative hydride indicates that nuclear dipole interaction works in spin transferring with a lesser role of atom exchange. In the solid film, not only the spin buildup but also the decay of enhanced polarization is faster than the thermal recovery rate for the bulk salt. Eliminating the signal of thick salt, we find that the nuclear spin polarization in the chloride film reaches over 100 times the thermal equilibrium.

Thermal radiative and thermodynamic properties of solid and liquid uranium and plutonium carbides in the visible-near-infrared range

NASA Astrophysics Data System (ADS)

Fisenko, Anatoliy I.; Lemberg, Vladimir F.

2016-09-01

The knowledge of thermal radiative and thermodynamic properties of uranium and plutonium carbides under extreme conditions is essential for designing a new metallic fuel materials for next generation of a nuclear reactor. The present work is devoted to the study of the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides at their melting/freezing temperatures. The Stefan-Boltzmann law, total energy density, number density of photons, Helmholtz free energy density, internal energy density, enthalpy density, entropy density, heat capacity at constant volume, pressure, and normal total emissivity are calculated using experimental data for the frequency dependence of the normal spectral emissivity of liquid and solid uranium and plutonium carbides in the visible-near infrared range. It is shown that the thermal radiative and thermodynamic functions of uranium carbide have a slight difference during liquid-to-solid transition. Unlike UC, such a difference between these functions have not been established for plutonium carbide. The calculated values for the normal total emissivity of uranium and plutonium carbides at their melting temperatures is in good agreement with experimental data. The obtained results allow to calculate the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides for any size of samples. Based on the model of Hagen-Rubens and the Wiedemann-Franz law, a new method to determine the thermal conductivity of metals and carbides at the melting points is proposed.

ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

NASA Technical Reports Server (NTRS)

Walton, J. T.

1994-01-01

ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

The nuclear thermal electric rocket: a proposed innovative propulsion concept for manned interplanetary missions

NASA Astrophysics Data System (ADS)

Dujarric, C.; Santovincenzo, A.; Summerer, L.

2013-03-01

Conventional propulsion technology (chemical and electric) currently limits the possibilities for human space exploration to the neighborhood of the Earth. If farther destinations (such as Mars) are to be reached with humans on board, a more capable interplanetary transfer engine featuring high thrust, high specific impulse is required. The source of energy which could in principle best meet these engine requirements is nuclear thermal. However, the nuclear thermal rocket technology is not yet ready for flight application. The development of new materials which is necessary for the nuclear core will require further testing on ground of full-scale nuclear rocket engines. Such testing is a powerful inhibitor to the nuclear rocket development, as the risks of nuclear contamination of the environment cannot be entirely avoided with current concepts. Alongside already further matured activities in the field of space nuclear power sources for generating on-board power, a low level investigation on nuclear propulsion has been running since long within ESA, and innovative concepts have already been proposed at an IAF conference in 1999 [1, 2]. Following a slow maturation process, a new concept was defined which was submitted to a concurrent design exercise in ESTEC in 2007. Great care was taken in the selection of the design parameters to ensure that this quite innovative concept would in all respects likely be feasible with margins. However, a thorough feasibility demonstration will require a more detailed design including the selection of appropriate materials and the verification that these can withstand the expected mechanical, thermal, and chemical environment. So far, the predefinition work made clear that, based on conservative technology assumptions, a specific impulse of 920 s could be obtained with a thrust of 110 kN. Despite the heavy engine dry mass, a preliminary mission analysis using conservative assumptions showed that the concept was reducing the required Initial Mass in Low Earth Orbit compared to conventional nuclear thermal rockets for a human mission to Mars. Of course, the realization of this concept still requires proper engineering and the dimensioning of quite unconventional machinery. A patent was filed on the concept. Because of the operating parameters of the nuclear core, which are very specific to this type of concept, it seems possible to test on ground this kind of engine at full scale in close loop using a reasonable size test facility with safe and clean conditions. Such tests can be conducted within fully confined enclosure, which would substantially increase the associated inherent nuclear safety levels. This breakthrough removes a showstopper for nuclear rocket engines development. The present paper will disclose the NTER (Nuclear Thermal Electric Rocket) engine concept, will present some of the results of the ESTEC concurrent engineering exercise, and will explain the concept for the NTER on-ground testing facility. Regulations and safety issues related to the development and implementation of the NTER concept will be addressed as well.

Nuclear fuels for very high temperature applications

NASA Astrophysics Data System (ADS)

Lundberg, L. B.; Hobbins, R. R.

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

IR Studies of the Spin-Nuclear Conversion in the Vicinity of alpha α - beta β - Transition in Cryodeposited Methane Films

NASA Astrophysics Data System (ADS)

Drobyshev, A.; Aldiyarov, A.; Sokolov, D.; Shinbayeva, A.

2017-06-01

Solid methane belongs to a group of crystals containing hydrogen atoms, whose macroscopic properties are greatly influenced by the spin interaction of hydrogen nuclei. In particular, the methane molecule, which has four protons with spin I=1/2, has three total spin modifications: para-, ortho- and meta-states with three values of the total spin moments of 0, 1 and 2, respectively. Equilibrium concentrations of these modifications and relaxation times are dependent on the temperature, affecting the observed thermal properties of solid methane, such as thermal conductivity, specific heat, thermal expansion. In this paper, we attempt to explain the peculiarities of thin film growth of methane at cryogenic temperatures from the viewpoint of spin-nuclear transformations. Our observations of absorption intensity at a frequency corresponding to 1/2 of the absorption band amplitude of deformation vibrations record a step-like change in the position of the absorption band during the sample deposition process. The observed phenomenon, in our opinion, is the demonstration of spin transformations during deposition.

Discrete Element Model for Simulations of Early-Life Thermal Fracturing Behaviors in Ceramic Nuclear Fuel Pellets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hai Huang; Ben Spencer; Jason Hales

2014-10-01

A discrete element Model (DEM) representation of coupled solid mechanics/fracturing and heat conduction processes has been developed and applied to explicitly simulate the random initiations and subsequent propagations of interacting thermal cracks in a ceramic nuclear fuel pellet during initial rise to power and during power cycles. The DEM model clearly predicts realistic early-life crack patterns including both radial cracks and circumferential cracks. Simulation results clearly demonstrate the formation of radial cracks during the initial power rise, and formation of circumferential cracks as the power is ramped down. In these simulations, additional early-life power cycles do not lead to themore » formation of new thermal cracks. They do, however clearly indicate changes in the apertures of thermal cracks during later power cycles due to thermal expansion and shrinkage. The number of radial cracks increases with increasing power, which is consistent with the experimental observations.« less

Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

NASA Astrophysics Data System (ADS)

Carcreff, H.; Salmon, L.; Lepeltier, V.; Guyot, J. M.; Bouard, E.

2018-01-01

Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the "zero method". Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the "zero method" measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between "zero" and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions, obtained with the Self-Power Neutron Detector suited to the CALMOS-2 calorimetric probe, are compared with those obtained with current devices. This campaign allowed to highlight advantages brought by the human machine interface automation, which deeply refined the profiles definition. Finally, the decay of the reactor residual power after shutdown could be performed after shutdown, demonstrating the ability of such type of calorimeter to follow the heating level whatever the thermohydraulic conditions, forced or natural convection regimes.

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

DOE PAGES

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

2016-11-18

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, V.

1984-06-13

This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

13C Nuclear magnetic resonance studies of kerogen from Cretaceous black shales thermally altered by basaltic intrusions and laboratory simulations

USGS Publications Warehouse

Dennis, L.W.; Maciel, G.E.; Hatcher, P.G.; Simoneit, B.R.T.

1982-01-01

Cretaceous black shales from DSDP Leg 41, Site 368 in the Eastern Atlantic Ocean were thermally altered during the Miocene by an intrusive basalt. The sediments overlying and underlying the intrusive body were subjected to high temperatures (up to ~ 500??C) and, as a result, their kerogen was significantly altered. The extent of this alteration has been determined by examination by means of 13C nuclear magnetic resonance, using cross polarization/magic-angle spinning (CP/MAS). Results indicate that the kerogen becomes progressively more aromatic in the vicinity of the intrusive body. Laboratory heating experiments, simulating the thermal effects of the basaltic intrusion, produced similar results on unaltered shale from the drill core. The 13C CP/MAS results appear to provide a good measure of thermal alteration. ?? 1982.

Melting phase relations in the Fe-S and Fe-S-O systems at core conditions in small terrestrial bodies

NASA Astrophysics Data System (ADS)

Pommier, Anne; Laurenz, Vera; Davies, Christopher J.; Frost, Daniel J.

2018-05-01

We report an experimental investigation of phase equilibria in the Fe-S and Fe-S-O systems. Experiments were performed at high temperatures (1400-1850 °C) and high pressures (14 and 20 GPa) using a multi-anvil apparatus. The results of this study are used to understand the effect of sulfur and oxygen on core dynamics in small terrestrial bodies. We observe that the formation of solid FeO grains occurs at the Fe-S liquid - Fe solid interface at high temperature ( > 1400 °C at 20 GPa). Oxygen fugacities calculated for each O-bearing sample show that redox conditions vary from ΔIW = -0.65 to 0. Considering the relative density of each phase and existing evolutionary models of terrestrial cores, we apply our experimental results to the cores of Mars and Ganymede. We suggest that the presence of FeO in small terrestrial bodies tends to contribute to outer-core compositional stratification. Depending on the redox and thermal history of the planet, FeO may also help form a transitional redox zone at the core-mantle boundary.

Robust techniques for polarization and detection of nuclear spin ensembles

NASA Astrophysics Data System (ADS)

Scheuer, Jochen; Schwartz, Ilai; Müller, Samuel; Chen, Qiong; Dhand, Ish; Plenio, Martin B.; Naydenov, Boris; Jelezko, Fedor

2017-11-01

Highly sensitive nuclear spin detection is crucial in many scientific areas including nuclear magnetic resonance spectroscopy, magnetic resonance imaging (MRI), and quantum computing. The tiny thermal nuclear spin polarization represents a major obstacle towards this goal which may be overcome by dynamic nuclear spin polarization (DNP) methods. The latter often rely on the transfer of the thermally polarized electron spins to nearby nuclear spins, which is limited by the Boltzmann distribution of the former. Here we utilize microwave dressed states to transfer the high (>92 % ) nonequilibrium electron spin polarization of a single nitrogen-vacancy center (NV) induced by short laser pulses to the surrounding 13C carbon nuclear spins. The NV is repeatedly repolarized optically, thus providing an effectively infinite polarization reservoir. A saturation of the polarization of the nearby nuclear spins is achieved, which is confirmed by the decay of the polarization transfer signal and shows an excellent agreement with theoretical simulations. Hereby we introduce the polarization readout by polarization inversion method as a quantitative magnetization measure of the nuclear spin bath, which allows us to observe by ensemble averaging macroscopically hidden polarization dynamics like Landau-Zener-Stückelberg oscillations. Moreover, we show that using the integrated solid effect both for single- and double-quantum transitions nuclear spin polarization can be achieved even when the static magnetic field is not aligned along the NV's crystal axis. This opens a path for the application of our DNP technique to spins in and outside of nanodiamonds, enabling their application as MRI tracers. Furthermore, the methods reported here can be applied to other solid state systems where a central electron spin is coupled to a nuclear spin bath, e.g., phosphor donors in silicon and color centers in silicon carbide.

Thermal conductivity measurements via time-domain thermoreflectance for the characterization of radiation induced damage

DOE PAGES

Cheaito, Ramez; Gorham, Caroline S.; Carnegie Mellon Univ., Pittsburgh, PA; ...

2015-05-01

The progressive build up of displacement damage and fission products inside different systems and components of a nuclear reactor can lead to significant defect formation, degradation, and damage of the constituent materials. This structural modification can highly influence the thermal transport mechanisms and various mechanical properties of solids. In this paper we demonstrate the use of time-domain thermoreflectance (TDTR), a non-destructive method capable of measuring the thermal transport in material systems from nano to bulk scales, to study the effect of radiation damage and the subsequent changes in the thermal properties of materials. We use TDTR to show that displacementmore » damage from ion irradiation can significantly reduce the thermal conductivity of Optimized ZIRLO, a material used as fuel cladding in several current nuclear reactors. We find that the thermal conductivity of copper-niobium nanostructured multilayers does not change with helium ion irradiation doses of up to 10 15 cm -2 and ion energy of 200 keV suggesting that these structures can be used and radiation tolerant materials in nuclear reactors. We compare the effect of ion doses and ion beam energies on the measured thermal conductivity of bulk silicon. Results demonstrate that TDTR thermal measurements can be used to quantify depth dependent damage.« less

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

Phase Equilibria of a S- and C-Poor Lunar Core

NASA Technical Reports Server (NTRS)

Righter, K.; Pando, K.; Go, B. M.; Danielson, L. R.; Habermann, M.

2016-01-01

The composition of the lunar core can have a large impact on its thermal evolution, possible early dynamo creation, and physical state. Geochemical measurements have placed better constraints on the S and C content of the lunar mantle. In this study we have carried out phase equilibrium studies of geochemically plausible S- and C-poor lunar core compositions in the Fe-Ni-S-C system, and apply them to the early history of the Moon. We chose two bulk core compositions, with differing S and C content based on geochemical analyses of S and C trapped melts in Apollo samples, and on the partitioning of S and C between metal and silicate. This approach allowed calculation of core S and C contents - 90% Fe, 9% Ni, 0.5% C, and 0.375% S by weight; a second composition contained 1% each of S and C. Experiments were carried out from 1473K to 1973K and 1 GPa to 5 GPa, in piston cylinder and multi- anvil apparatuses. Combination of the thermal model of with our results, shows that a solid inner core (and therefore initiation of a dynamo) may have been possible in the earliest history of the Moon (approximately 4.2 Ga ago), in agreement with. Thus a volatile poor lunar core may explain the thermal and magnetic history of the Moon.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Thermal-work strain in law enforcement personnel during chemical, biological, radiological, and nuclear (CBRN) training

PubMed Central

Yokota, M; Karis, A J; Tharion, W J

2014-01-01

Background: Thermal safety standards for the use of chemical, biological, radiological, and nuclear (CBRN) ensembles have been established for various US occupations, but not for law enforcement personnel. Objectives: We examined thermal strain levels of 30 male US law enforcement personnel who participated in CBRN field training in Arizona, Florida, and Massachusetts. Methods: Physiological responses were examined using unobtrusive heart rate (HR) monitors and a simple thermoregulatory model to predict core temperature (Tc) using HR and environment. Results: Thermal strain levels varied by environments, activity levels, and type of CBRN ensemble. Arizona and Florida volunteers working in hot-dry and hot-humid environment indicated high heat strain (predicted max Tc>38.5°C). The cool environment of Massachusetts reduced thermal strain although thermal strains were occasionally moderate. Conclusions: The non-invasive method of using physiological monitoring and thermoregulatory modeling could improve law enforcement mission to reduce the risk of heat illness or injury. PMID:24999847

Effects of hydrogen on thermal creep behaviour of Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

Suman, Siddharth; Khan, Mohd Kaleem; Pathak, Manabendra; Singh, R. N.

2018-01-01

Zirconium alloys are extensively used for nuclear fuel cladding. Creep is one of the most likely degradation mechanisms for fuel cladding during reactor operating and repository conditions. Fuel cladding tubes undergo waterside corrosion during service and hydrogen is produced as a result of it-a fraction of which is picked up by cladding. Hydrogen remains in solid solution up to terminal solid solubility and it precipitates as brittle hydride phase in the zirconium metal matrix beyond this limiting concentration. Hydrogen, either in solid solution or as precipitated hydride, alters the creep behaviour of zirconium alloys. The present article critically reviews the influence of hydrogen on thermal creep behaviour of zirconium alloys, develops the systematic understanding of this multifaceted phenomenon, and delineates the thrust areas which require further investigations.

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

NUCLEAR REACTOR CORE DESIGN

DOEpatents

Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

1960-03-22

An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

On a thermal analysis of a second stripper for rare isotope accelerator.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Momozaki, Y.; Nolen, J.; Nuclear Engineering Division

2008-08-04

This memo summarizes simple calculations and results of the thermal analysis on the second stripper to be used in the driver linac of Rare Isotope Accelerator (RIA). Both liquid (Sodium) and solid (Titanium and Vanadium) stripper concepts were considered. These calculations were intended to provide basic information to evaluate the feasibility of liquid (thick film) and solid (rotating wheel) second strippers. Nuclear physics calculations to estimate the volumetric heat generation in the stripper material were performed by 'LISE for Excel'. In the thermal calculations, the strippers were modeled as a thin 2D plate with uniform heat generation within the beammore » spot. Then, temperature distributions were computed by assuming that the heat spreads conductively in the plate in radial direction without radiative heat losses to surroundings.« less

Mechatronic Materials and Systems. Design and Demonstration of High Aughtority Shape Morphing Structures

DTIC Science & Technology

2005-09-01

thermal expansion of these truss elements. One side of the structure is fully clamped, while the other is free to displace. As in prior assessments [6...levels, by using the finite element package ABAQUS . To simulate the complete system, the core and the Kagome face members are modeled using linear...code ABAQUS . To simulate the complete actuation system, the core and Kagome members are modeled using linear Timoshenko-type beams, while the solid

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Worrall, Andrew; Todosow, Michael

2016-01-01

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include:more » increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated« less

Phase equilibria of a low S and C lunar core: Implications for an early lunar dynamo and physical state of the current core

NASA Astrophysics Data System (ADS)

Righter, K.; Go, B. M.; Pando, K. A.; Danielson, L.; Ross, D. K.; Rahman, Z.; Keller, L. P.

2017-04-01

Multiple lines of geochemical and geophysical evidence suggest the Moon has a small metallic core, yet the composition of the core is poorly constrained. The physical state of the core (now or in the past) depends on detailed knowledge of its composition, and unfortunately, there is little available data on relevant multicomponent systems (i.e., Fe-Ni-S-C) at lunar interior conditions. In particular, there is a dearth of phase equilibrium data to elucidate whether a specific core composition could help to explain an early lunar geodynamo and magnetic field intensities, or current solid inner core/liquid outer core states. We utilize geochemical information to estimate the Ni, S and C contents of the lunar core, and then carry out phase equilibria experiments on several possible core compositions at the pressure and temperature conditions relevant to the lunar interior. The first composition is 0.5 wt% S and 0.375 wt% C, based on S and C contents of Apollo glasses. A second composition contains 1 wt% each of S and C, and assumes that the lunar mantle experienced degassing of up to 50% of its S and C. Finally a third composition contains C as the dominant light element. Phase equilibrium experiments were completed at 1, 3 and 5 GPa, using piston cylinder and multi-anvil techniques. The first composition has a liquidus near 1550 °C and solidus near 1250 °C. The second composition has a narrower liquidus and solidus temperatures of 1400 and 1270 °C, respectively, while the third composition is molten down to 1150 °C. As the composition crystallizes, the residual liquid becomes enriched in S and C, but S enrichment is greater due to the incorporation of C (but not S) into solid metallic FeNi. Comparison of these results to thermal models for the Moon allow an evaluation of which composition is consistent with the geophysical data of an early dynamo and a currently solid inner and liquid outer core. Composition 1 has a high enough liquidus to start crystallizing early in lunar history (4.3 Ga), consistent with the possible core dynamo initiated by crystallization of a solid inner core. Composition 1 also stays partially molten throughout lunar history, and could easily explain the seismic data. Composition 2, on the other hand, can satisfy one or the other set of geophysical data, but not both and thus seems like a poor candidate for a lunar core composition. Composition 3 remains molten to temperatures that are lower than current estimates for the lunar core, thus ruling out the possibility of a C-rich (and S-poor) lunar core. The S- and C-poor core composition studied here (composition 1) is consistent with all available geochemical and geophysical data and provides a simple heat source and mechanism for a lunar core dynamo (core crystallization) that would obviate the need for other primary mechanisms such as impacts, core-mantle coupling, or unusual thermal histories.

Time-dependent heat transfer in the spherical Earth: Implications on the power and thermal evolution of the core

NASA Astrophysics Data System (ADS)

Hofmeister, A. M.; Criss, R. E.

2015-12-01

We quantitatively investigate the time-dependence of heat conduction for a post-core, spherical Earth that is not convecting, due to compositional layering, based on hundreds of measurements of thermal diffusivity (D) for insulators and metals. Consistency of our solutions for widely ranging input parameters indicates how additional heat transfer mechanisms (mantle magmatism and convection) affect thermal evolution of the core. We consider 1) interior starting temperatures (T) of 273-5000 K, which represent variations in primordial heat, 2) different distributions and decay of long-lived radioactive isotopes, 3) additional heat sources in the core (primordial or latent heat), and 4) variable depth-T dependence of D. Our new analytical solution for cooling of a constant D sphere validates our numerical results. The bottom line is that the thermally insulating nature of minerals, combined with constraints of spherical geometry, limits steep thermal gradients to the upper mantle, consistent with the short length scale (x ~700 km) of cooling over t = 4.5 Ga indicated by dimensional analysis [x2 ~ 4Dt], and with plate tectonics. Consequently, interior temperatures vary little so the core has remained hot and is possibly warming. Findings include: 1) Constant vs. variable D affects thermal profiles only in detail, with D for the metallic core being inconsequential. 2) The hottest zone in Earth may lie in the uppermost lower mantle; 3) Most radiogenic heat is released in Earth's outermost 1000 km thereby driving an active outer shell; 4) Earth's core is essentially isothermal and is thus best described by the liquid-solid phase boundary; 5) Deeply sequestered radioactivity or other heat will melt the core rather than by run the dynamo (note that the heat needed to have melted the outer core is 10% of radiogenic heat generated over Earth's history); 6) Inefficient cooling of an Earth-sized mass means that heat essentially remains where it is generated, until it is removed by magmatism; 7) Importantly, the observed plate velocities are consistent with a Nusselt number of 1, i.e. the present day cooling is essentially conductive. Conductive cooling plus magmatism largely governs Earth's thermal structure and dynamics, below a unicellular upper mantle. Core dynamics and magnetism are likely driven by rotational effects.

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szilard, Ronaldo Henriques

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

Ceramic Borehole Seals for Nuclear Waste Disposal Applications

NASA Astrophysics Data System (ADS)

Lowry, B.; Coates, K.; Wohletz, K.; Dunn, S.; Patera, E.; Duguid, A.; Arnold, B.; Zyvoloski, G.; Groven, L.; Kuramyssova, K.

2015-12-01

Sealing plugs are critical features of the deep borehole system design. They serve as structural platforms to bear the weight of the backfill column, and as seals through their low fluid permeability and bond to the borehole or casing wall. High hydrostatic and lithostatic pressures, high mineral content water, and elevated temperature due to the waste packages and geothermal gradient challenge the long term performance of seal materials. Deep borehole nuclear waste disposal faces the added requirement of assuring performance for thousands of years in large boreholes, requiring very long term chemical and physical stability. A high performance plug system is being developed which capitalizes on the energy of solid phase reactions to form a ceramic plug in-situ. Thermites are a family of self-oxidized metal/oxide reactions with very high energy content and the ability to react under water. When combined with engineered additives the product exhibits attractive structural, sealing, and corrosion properties. In the initial phase of this research, exploratory and scaled tests demonstrated formulations that achieved controlled, fine grained, homogeneous, net shape plugs composed predominantly of ceramic material. Laboratory experiments produced plug cores with confined fluid permeability as low as 100 mDarcy, compressive strength as high as 70 MPa (three times the strength of conventional well cement), with the inherent corrosion resistance and service temperature of ceramic matrices. Numerical thermal and thermal/structural analyses predicted the in-situ thermal performance of the reacted plugs, showing that they cooled to ambient temperature (and design strength) within 24 to 48 hours. The current development effort is refining the reactant formulations to achieve desired performance characteristics, developing the system design and emplacement processes to be compatible with conventional well service practices, and understanding the thermal, fluid, and structural effects the plug will have on surrounding media. This paper will report on the state of the development effort and plans for a field demonstration in early 2016 in a cased well with traditional plug seal and strength measurements.

An End-To-End Test of A Simulated Nuclear Electric Propulsion System

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Hrbud, Ivana; Goddfellow, Keith; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

The Safe Affordable Fission Engine (SAFE) test series addresses Phase I Space Fission Systems issues in it particular non-nuclear testing and system integration issues leading to the testing and non-nuclear demonstration of a 400-kW fully integrated flight unit. The first part of the SAFE 30 test series demonstrated operation of the simulated nuclear core and heat pipe system. Experimental data acquired in a number of different test scenarios will validate existing computational models, demonstrated system flexibility (fast start-ups, multiple start-ups/shut downs), simulate predictable failure modes and operating environments. The objective of the second part is to demonstrate an integrated propulsion system consisting of a core, conversion system and a thruster where the system converts thermal heat into jet power. This end-to-end system demonstration sets a precedent for ground testing of nuclear electric propulsion systems. The paper describes the SAFE 30 end-to-end system demonstration and its subsystems.

Role of nuclear grade graphite in controlling oxidation in modular HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, Willaim; Strydom, G.; Kane, J.

2014-11-01

The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of coremore » environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.« less

Novel Scintillating Materials Based on Phenyl-Polysiloxane for Neutron Detection and Monitoring

NASA Astrophysics Data System (ADS)

Degerlier, M.; Carturan, S.; Gramegna, F.; Marchi, T.; Palma, M. Dalla; Cinausero, M.; Maggioni, G.; Quaranta, A.; Collazuol, G.; Bermudez, J.

Neutron detectors are extensively used at many nuclear research facilities across Europe. Their application range covers many topics in basic and applied nuclear research: in nuclear structure and reaction dynamics (reaction reconstruction and decay studies); in nuclear astrophysics (neutron emission probabilities); in nuclear technology (nuclear data measurements and in-core/off-core monitors); in nuclear medicine (radiation monitors, dosimeters); in materials science (neutron imaging techniques); in homeland security applications (fissile materials investigation and cargo inspection). Liquid scintillators, widely used at present, have however some drawbacks given by toxicity, flammability, volatility and sensitivity to oxygen that limit their duration and quality. Even plastic scintillators are not satisfactory because they have low radiation hardness and low thermal stability. Moreover organic solvents may affect their optical properties due to crazing. In order to overcome these problems, phenyl-polysiloxane based scintillators have been recently developed at Legnaro National Laboratory. This new solution showed very good chemical and thermal stability and high radiation hardness. The results on the different samples performance will be presented, paying special attention to a characterization comparison between synthesized phenyl containing polysiloxane resins where a Pt catalyst has been used and a scintillating material obtained by condensation reaction, where tin based compounds are used as catalysts. Different structural arrangements as a result of different substituents on the main chain have been investigated by High Resolution X-Ray Diffraction, while the effect of improved optical transmittance on the scintillation yield has been elucidated by a combination of excitation/fluorescence measurements and scintillation yield under exposure to alpha and γ-rays.

Thermal oxidation of nuclear graphite: A large scale waste treatment option.

PubMed

Theodosiou, Alex; Jones, Abbie N; Marsden, Barry J

2017-01-01

This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400-1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700-800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000-1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput.

Thermal oxidation of nuclear graphite: A large scale waste treatment option

PubMed Central

Jones, Abbie N.; Marsden, Barry J.

2017-01-01

This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400–1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700–800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000–1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput. PMID:28793326

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Propulsion Systems Panel deliberations

NASA Technical Reports Server (NTRS)

Bianca, Carmelo J.; Miner, Robert; Johnston, Lawrence M.; Bruce, R.; Dennies, Daniel P.; Dickenson, W.; Dreshfield, Robert; Karakulko, Walt; Mcgaw, Mike; Munafo, Paul M.

1993-01-01

The Propulsion Systems Panel was established because of the specialized nature of many of the materials and structures technology issues related to propulsion systems. This panel was co-chaired by Carmelo Bianca, MSFC, and Bob Miner, LeRC. Because of the diverse range of missions anticipated for the Space Transportation program, three distinct propulsion system types were identified in the workshop planning process: liquid propulsion systems, solid propulsion systems and nuclear electric/nuclear thermal propulsion systems.

Propulsion Systems Panel deliberations

NASA Astrophysics Data System (ADS)

Bianca, Carmelo J.; Miner, Robert; Johnston, Lawrence M.; Bruce, R.; Dennies, Daniel P.; Dickenson, W.; Dreshfield, Robert; Karakulko, Walt; McGaw, Mike; Munafo, Paul M.

1993-02-01

The Propulsion Systems Panel was established because of the specialized nature of many of the materials and structures technology issues related to propulsion systems. This panel was co-chaired by Carmelo Bianca, MSFC, and Bob Miner, LeRC. Because of the diverse range of missions anticipated for the Space Transportation program, three distinct propulsion system types were identified in the workshop planning process: liquid propulsion systems, solid propulsion systems and nuclear electric/nuclear thermal propulsion systems.

Melting relations in the iron-sulfur system at ultra-high pressures - Implications for the thermal state of the earth

NASA Technical Reports Server (NTRS)

Williams, Quentin; Jeanloz, Raymond

1990-01-01

The melting temperatures of FeS-troilite and of a 10-wt-pct sulfur iron alloy have been measured to pressures of 120 and 90 GPa, respectively. The results document that FeS melts at a temperature of 4100 (+ or - 300) K at the pressure of the core-mantle boundary. Eutecticlike behavior persists in the iron-sulfur system to the highest pressures of measurements, in marked contrast to the solid-solutionlike behavior observed at high pressures in the iron-iron oxide system. Iron with 10-wt-pct sulfur melts at a similar temperature as FeS at core-mantle boundary conditions. If the sole alloying elements of iron within the core are sulfur and oxygen and the outer core is entirely liquid, the minimum temperature at the top of the outer core is 4900 (+ or - 400) K. Calculations of mantle geotherms dictate that there must be a temperature increase of between 1000 and 2000 K across thermal boundary layers within the mantle. If D-double-prime is compositionally stratified, it could accommodate the bulk of this temperature jump.

Fiber Bragg Gratings for High-Temperature Thermal Characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stinson-Bagby, Kelly L.; Fielder, Robert S.

2004-07-01

Fiber Bragg grating (FBG) sensors were used as a characterization tool to study the SAFE-100 thermal simulator at the Nasa Marshal Space Flight Center. The motivation for this work was to support Nasa space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. Distributed high temperature measurements, up to 1150 deg. C, were made with FBG temperature sensors. Additionally, FBG strain measurements were taken at elevated temperatures to provide a strain profile of the core during operation. This paper will discuss the contribution of these measurements to meet the goals of Nasa Marshallmore » Space Flight Center's Propulsion Research Center. (authors)« less

Time-dependent simulations of disk-embedded planetary atmospheres

NASA Astrophysics Data System (ADS)

Stökl, A.; Dorfi, E. A.

2014-03-01

At the early stages of evolution of planetary systems, young Earth-like planets still embedded in the protoplanetary disk accumulate disk gas gravitationally into planetary atmospheres. The established way to study such atmospheres are hydrostatic models, even though in many cases the assumption of stationarity is unlikely to be fulfilled. Furthermore, such models rely on the specification of a planetary luminosity, attributed to a continuous, highly uncertain accretion of planetesimals onto the surface of the solid core. We present for the first time time-dependent, dynamic simulations of the accretion of nebula gas into an atmosphere around a proto-planet and the evolution of such embedded atmospheres while integrating the thermal energy budget of the solid core. The spherical symmetric models computed with the TAPIR-Code (short for The adaptive, implicit RHD-Code) range from the surface of the rocky core up to the Hill radius where the surrounding protoplanetary disk provides the boundary conditions. The TAPIR-Code includes the hydrodynamics equations, gray radiative transport and convective energy transport. The results indicate that diskembedded planetary atmospheres evolve along comparatively simple outlines and in particular settle, dependent on the mass of the solid core, at characteristic surface temperatures and planetary luminosities, quite independent on numerical parameters and initial conditions. For sufficiently massive cores, this evolution ultimately also leads to runaway accretion and the formation of a gas planet.

Solubility of K in Fe-S liquid, silicate-K/Fe-S/liq equilibria, and their planetary implications

NASA Technical Reports Server (NTRS)

Gangully, J.; Kennedy, G. C.

1977-01-01

Potassium has been found to have extremely limited absolute solubility in Fe-S liquid in the pressure-temperature range of 18 to 40 kbars, 1050 to 1150 C, and fO2 within the field of metallic iron. It also partitioned into a certain silicate phase highly in preference to Fe-S liquid at 30 kbar and 1100 C. The dependence of the partitioning of K between solid silicate and Fe-S liquid on fO2 and compositions of mineral solid solutions have been analyzed. These experimental data, along with those of others, limit the amount of K that could fractionate in Fe-S liquid layers or a core in the early history of the moon and, thus, act as localized heat sources in its thermal history models; the data also seem to argue against a chondritic abundance of potassium for earth. The question of fractionation of enough K-40 in an Fe-S liquid outer core of earth to provide the necesary thermal energy for the geomagnetic dynamo remains unresolved.

New insights into the thermal behaviour of organic ionic plastic crystals: magnetic resonance imaging of polycrystalline morphology alterations induced by solid-solid phase transitions.

PubMed

Romanenko, Konstantin; Pringle, Jennifer M; O'Dell, Luke A; Forsyth, Maria

2015-07-15

Organic ionic plastic crystals (OIPCs) show strong potential as solid-state electrolytes for lithium battery applications, demonstrating promising electrochemical performance and eliminating the need for a volatile and flammable liquid electrolyte. The ionic conductivity (σ) in these systems has recently been shown to depend strongly on polycrystalline morphology, which is largely determined by the sample's thermal history. [K. Romanenko et al., J. Am. Chem. Soc., 2014, 136, 15638]. Tailoring this morphology could lead to conductivities sufficiently high for battery applications, so a more complete understanding of how phenomena such as solid-solid phase transitions can affect the sample morphology is of significant interest. Anisotropic relaxation of nuclear spin magnetisation provides a new MRI based approach for studies of polycrystalline materials at both a macroscopic and molecular level. In this contribution, morphology alterations induced by solid-solid phase transitions in triisobutyl(methyl)phosphonium bis(fluorosulfonyl)imide (P1444FSI) and diethyl(methyl)(isobutyl)phosphonium hexafluorophosphate (P1224PF6) are examined using magnetic resonance imaging (MRI), alongside nuclear magnetic resonance (NMR) spectroscopy, diffusion measurements and conductivity data. These observations are linked to molecular dynamics and structural behaviour crucial for the conductive properties of OIPCs. A distinct correlation is established between the conductivity at a given temperature, σ(T), and the intensity of the narrow NMR signal that is attributed to a mobile fraction, fm(T), of ions in the OIPC. To explain these findings we propose an analogy with the well-studied relationship between permeability (k) and void fraction (θ) in porous media, with k(θ) commonly quantified by a power-law dependence that can also be employed to describe σ(fm).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density andmore » inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.« less

Non-nuclear Testing of Reactor Systems in the Early Flight Fission Test Facilities (EFF-TF)

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2004-01-01

The Early Flight Fission-Test Facility (EFF-TF) can assist in the &sign and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are "non-nuclear" in nature (e.g. system s nuclear operations are understood). For many systems. thermal simulators can he used to closely mimic fission heat deposition. Axial power profile, radial power profile. and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other NASA centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004.

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Nuclear Thermal Rocket/Vehicle Design Options for Future NASA Missions to the Moon and Mars

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Corban, Robert R.; Mcguire, Melissa L.; Beke, Erik G.

1995-01-01

The nuclear thermal rocket (NTR) provides a unique propulsion capability to planners/designers of future human exploration missions to the Moon and Mars. In addition to its high specific impulse (approximately 850-1000 s) and engine thrust-to-weight ratio (approximately 3-10), the NTR can also be configured as a 'dual mode' system capable of generating electrical power for spacecraft environmental systems, communications, and enhanced stage operations (e.g., refrigeration for long-term liquid hydrogen storage). At present the Nuclear Propulsion Office (NPO) is examining a variety of mission applications for the NTR ranging from an expendable, single-burn, trans-lunar injection (TLI) stage for NASA's First Lunar Outpost (FLO) mission to all propulsive, multiburn, NTR-powered spacecraft supporting a 'split cargo-piloted sprint' Mars mission architecture. Each application results in a particular set of requirements in areas such as the number of engines and their respective thrust levels, restart capability, fuel operating temperature and lifetime, cryofluid storage, and stage size. Two solid core NTR concepts are examined -- one based on NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor (NDR) technology, and a second concept which utilizes a ternary carbide 'twisted ribbon' fuel form developed by the Commonwealth of Independent States (CIS). The NDR and CIS concepts have an established technology database involving significant nuclear testing at or near representative operating conditions. Integrated systems and mission studies indicate that clusters of two to four 15 to 25 klbf NDR or CIS engines are sufficient for most of the lunar and Mars mission scenarios currently under consideration. This paper provides descriptions and performance characteristics for the NDR and CIS concepts, summarizes NASA's First Lunar Outpost and Mars mission scenarios, and describes characteristics for representative cargo and piloted vehicles compatible with a reference 240 t-class heavy lift launch vehicle (HLLV) and smaller 120 t HLLV option. Attractive performance characteristics and high-leverage technologies associated with both the engine and stage are identified, and supporting parametric sensitivity data is provided. The potential for commonality of engine and stage components to satisfy a broad range of lunar and Mars missions is also discussed.

Atmospheric Mining in the Outer Solar System: Aerial Vehicle Mission and Design Issues

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2015-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and deuterium can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and deuterium were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate, storage options, and different methods of direct use of the captured gases. While capturing 3He, large amounts of hydrogen and 4He are produced. With these two additional gases, the potential for fueling small and large fleets of additional exploration and exploitation vehicles exists. The mining aerospacecraft (ASC) could fly through the outer planet atmospheres, for global weather observations, localized storm or other disturbance investigations, wind speed measurements, polar observations, etc. Analyses of orbital transfer vehicles (OTVs), landers, and in-situ resource utilization (ISRU) mining factories are included. Preliminary observations are presented on near-optimal selections of moon base orbital locations, OTV power levels, and OTV and lander rendezvous points.

Re-interpretation of the ERMINE-V experiment validation of fission product integral cross section in the fast energy range

NASA Astrophysics Data System (ADS)

Ros, Paul; Leconte, Pierre; Blaise, Patrick; Naymeh, Laurent

2017-09-01

The current knowledge of nuclear data in the fast neutron energy range is not as good as in the thermal range, resulting in larger propagated uncertainties in integral quantities such as critical masses or reactivity effects. This situation makes it difficult to get the full benefit from recent advances in modeling and simulation. Zero power facilities such as the French ZPR MINERVE have already demonstrated that they can contribute to significantly reduce those uncertainties thanks to dedicated experiments. Historically, MINERVE has been mainly dedicated to thermal spectrum studies. However, experiments involving fast-thermal coupled cores were also performed in MINERVE as part of the ERMINE program, in order to improve nuclear data in fast spectra for the two French SFRs: PHENIX and SUPERPHENIX. Some of those experiments have been recently revisited. In particular, a full characterization of ZONA-1 and ZONA-3, two different cores loaded in the ERMINE V campaign, has been done, with much attention paid to possible sources of errors. It includes detailed geometric descriptions, energy profiles of the direct and adjoint fluxes and spectral indices obtained thanks to Monte Carlo calculations and compared to a reference fast core configuration. Sample oscillation experiments of separated fission products such as 103Rh or 99Tc, which were part of the ERMINE V program, have been simulated using recently-developed options in the TRIPOLI-4 code and compared to the experimental values. The present paper describes the corresponding results. The findings motivate in-depth studies for designing optimized coupled-core conditions in ZEPHYR, a new ZPR which will replace MINERVE and will provide integral data to meet the needs of Gen-III and Gen-IV reactors.

Summary of comparison and analysis of results from exercises 1 and 2 of the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mkhabela, P.; Han, J.; Tyobeka, B.

2006-07-01

The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor 400 MW design (PBMR-400) coupled neutronics/thermal hydraulics transient benchmark problem as part of their official activities. The scope of the benchmark is to establish a well-defined problem, based on a common given library of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark includes three steady state exercises andmore » six transient exercises. This paper describes the first two steady state exercises, their objectives and the international participation in terms of organization, country and computer code utilized. This description is followed by a comparison and analysis of the participants' results submitted for these two exercises. The comparison of results from different codes allows for an assessment of the sensitivity of a result to the method employed and can thus help to focus the development efforts on the most critical areas. The two first exercises also allow for removing of user-related modeling errors and prepare core neutronics and thermal-hydraulics models of the different codes for the rest of the exercises in the benchmark. (authors)« less

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

NASA Astrophysics Data System (ADS)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

Nuclear magnetic resonance study of thermal oxidation of polyisoprene

NASA Technical Reports Server (NTRS)

Golub, M. A.; Hsu, M. S.

1975-01-01

An investigation was conducted concerning the microstructural changes occurring in cis- and trans-1,4-polyisoprenes during uncatalized thermal oxidation in the solid phase. The investigation made use of approaches based on proton and carbon-13 NMR spectroscopy. The oxidation of squalene and dihydromyrcene in the liquid phase was also studied. The studies provide the first NMR spectroscopic evidence for the presence of epoxy and peroxide, hydroperoxide, and alcohol groups within the oxidized polyisoprene chain.

Preparation of a deuterated polymer: Simulating to produce a solid tritium radioactive source

NASA Astrophysics Data System (ADS)

Hu, Rui; Kan, Wentao; Xiong, Xiaoling; Wei, Hongyuan

2017-08-01

The preparation of a deuterated polymer was performed in order to simulate the production of the corresponding tritiated polymer as a solid tritium radioactive source. Substitution and addition reaction were used to introduce deuterium into the polymer. Proton nuclear magnetic resonance and FT-IR spectroscopy were used to investigate the extent and location of deuterium in the polymer, indicating an effectively deuterated polymer was produced. The thermal analysis showed that the final polymer product could tolerate the environmental temperature below 125 °C in its application. This research provides a prosperous method to prepare solid tritium radioactive source.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

End-to-End Demonstrator of the Safe Affordable Fission Engine (SAFE) 30: Power Conversion and Ion Engine Operation

NASA Technical Reports Server (NTRS)

Hrbud, Ivana; VanDyke, Melissa; Houts, Mike; Goodfellow, Keith; Schafer, Charles (Technical Monitor)

2001-01-01

The Safe Affordable Fission Engine (SAFE) test series addresses Phase 1 Space Fission Systems issues in particular non-nuclear testing and system integration issues leading to the testing and non-nuclear demonstration of a 400-kW fully integrated flight unit. The first part of the SAFE 30 test series demonstrated operation of the simulated nuclear core and heat pipe system. Experimental data acquired in a number of different test scenarios will validate existing computational models, demonstrated system flexibility (fast start-ups, multiple start-ups/shut downs), simulate predictable failure modes and operating environments. The objective of the second part is to demonstrate an integrated propulsion system consisting of a core, conversion system and a thruster where the system converts thermal heat into jet power. This end-to-end system demonstration sets a precedent for ground testing of nuclear electric propulsion systems. The paper describes the SAFE 30 end-to-end system demonstration and its subsystems.

Inhibition of ovarian cancer cell proliferation by a cell cycle inhibitory peptide fused to a thermally responsive polypeptide carrier.

PubMed

Massodi, Iqbal; Moktan, Shama; Rawat, Aruna; Bidwell, Gene L; Raucher, Drazen

2010-01-15

Current treatment of solid tumors is limited by normal tissue tolerance, resulting in a narrow therapeutic index. To increase drug specificity and efficacy and to reduce toxicity in normal tissues, we have developed a polypeptide carrier for a cell cycle inhibitory peptide, which has the potential to be thermally targeted to the tumor site. The design of this polypeptide is based on elastin-like polypeptide (ELP). The coding sequence of ELP was modified by the addition of the cell penetrating peptide Bac-7 at the N-terminus and a 23 amino acid peptide derived from p21 at the C-terminus (Bac-ELP1-p21). Bac-ELP1-p21 is soluble in aqueous solutions below physiological temperature (37 degrees C) but aggregates when the temperature is raised above 39 degrees C, making it a promising thermally responsive therapeutic carrier that may be actively targeted to solid tumors by application of focused hyperthermia. While Bac-ELP1-p21 at 37 degrees C did not have any effect on SKOV-3 cell proliferation, the use of hyperthermia increased the antiproliferative effect of Bac-ELP1-p21 compared with a thermally unresponsive control polypeptide. Bac-ELP1-p21 displayed both a cytoplasmic and nuclear distribution in the SKOV-3 cells, with nuclear-localized polypeptide enriched in the heated cells, as revealed by confocal microscopy. Using Western blotting, we show that Bac-ELP1-p21 caused a decrease in Rb phosphorylation levels in cells treated at 42 degrees C. The polypeptide also induced caspase activation, PARP cleavage, and cell cycle arrest in S-phase and G2/M-phase. These studies indicate that ELP is a promising macromolecular carrier for the delivery of cell cycle inhibitory peptides to solid tumors.

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

NASA Astrophysics Data System (ADS)

Knight, Travis W.; Anghaie, Samim

2002-11-01

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

Long-period seismology on Europa: 1. Physically consistent interior models

NASA Astrophysics Data System (ADS)

Cammarano, F.; Lekic, V.; Manga, M.; Panning, M.; Romanowicz, B.

2006-12-01

In order to examine the potential of seismology to determine the interior structure and properties of Europa, it is essential to calculate seismic velocities and attenuation for the range of plausible interiors. We calculate a range of models for the physical structure of Europa, as constrained by the satellite's composition, mass, and moment of inertia. We assume a water-ice shell, a pyrolitic or a chondritic mantle, and a core composed of pure iron or iron plus 20 weight percent of sulfur. We consider two extreme mantle thermal states: hot and cold. Given a temperature and composition, we determine density, seismic velocities, and attenuation using thermodynamical models. While anelastic effects will be negligible in a cold mantle and the brittle part of the ice shell, strong dispersion and dissipation are expected in a hot convective mantle and the bulk of the ice shell. There is a strong relationship between different thermal structures and compositions. The ``hot'' mantle may maintain temperatures consistent with a liquid core made of iron plus light elements. For the ``cold scenarios,'' the possibility of a solid iron core cannot be excluded, and it may even be favored. The depths of the ocean and core-mantle boundary are determined with high precision, 10 km and 40 km, respectively, once we assume a composition and thermal structure. Furthermore, the depth of the ocean is relatively insensitive (4 km) to the core composition used.

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

DOEpatents

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems; NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) QUARTERLY PROGRESS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pavel Hejzlar, Peter Yarsky, Mike Driscoll, Dan Wachs, Kevan Weaver, Ken Czerwinski, Mike Pope, James Parry, Theron D. Marshall, Cliff B. Davis, Dustin Crawford, Thomas Hartmann, Pradip Saha; Hejzlar, Pavel; Yarsky, Peter

2005-01-31

This project is organized under four major tasks (each of which has two or more subtasks) with contributions among the three collaborating organizations (MIT, INEEL and ANL-West): Task A: Core Physics and Fuel Cycle; Task B: Core Thermal Hydraulics; Task C: Plant Design; Task D: Fuel Design The lead PI, Michael J. Driscoll, has consolidated and summarized the technical progress submissions provided by the contributing investigators from all sites, under the above principal task headings.

COUPLED FAST-THERMAL POWER BREEDER REACTOR

DOEpatents

Avery, R.

1961-07-18

A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

Effective Thermal Property Estimation of Unitary Pebble Beds Based on a CFD-DEM Coupled Method for a Fusion Blanket

NASA Astrophysics Data System (ADS)

Chen, Lei; Chen, Youhua; Huang, Kai; Liu, Songlin

2015-12-01

Lithium ceramic pebble beds have been considered in the solid blanket design for fusion reactors. To characterize the fusion solid blanket thermal performance, studies of the effective thermal properties, i.e. the effective thermal conductivity and heat transfer coefficient, of the pebble beds are necessary. In this paper, a 3D computational fluid dynamics discrete element method (CFD-DEM) coupled numerical model was proposed to simulate heat transfer and thereby estimate the effective thermal properties. The DEM was applied to produce a geometric topology of a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. Based on this geometric topology, a CFD model was built to analyze the temperature distribution and obtain the effective thermal properties. The current numerical model was shown to be in good agreement with the existing experimental data for effective thermal conductivity available in the literature. supported by National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2015GB108002, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

Thermal Characterization of a Simulated Fission Engine via Distributed Fiber Bragg Gratings

NASA Astrophysics Data System (ADS)

Duncan, Roger G.; Fielder, Robert S.; Seeley, Ryan J.; Kozikowski, Carrie L.; Raum, Matthew T.

2005-02-01

We report the use of distributed fiber Bragg gratings to monitor thermal conditions within a simulated nuclear reactor core located at the Early Flight Fission Test Facility of the NASA Marshall Space Flight Center. Distributed fiber-optic temperature measurements promise to add significant capability and advance the state-of-the-art in high-temperature sensing. For the work reported herein, seven probes were constructed with ten sensors each for a total of 70 sensor locations throughout the core. These discrete temperature sensors were monitored over a nine hour period while the test article was heated to over 700 °C and cooled to ambient through two operational cycles. The sensor density available permits a significantly elevated understanding of thermal effects within the simulated reactor. Fiber-optic sensor performance is shown to compare very favorably with co-located thermocouples where such co-location was feasible.

The role of electronic energy loss in ion beam modification of materials

DOE PAGES

Weber, William J.; Duffy, Dorothy M.; Thome, Lionel; ...

2014-10-05

The interaction of energetic ions with solids results in energy loss to both atomic nuclei and electrons in the solid. In this article, recent advances in understanding and modeling the additive and competitive effects of nuclear and electronic energy loss on the response of materials to ion irradiation are reviewed. Experimental methods and large-scale atomistic simulations are used to study the separate and combined effects of nuclear and electronic energy loss on ion beam modification of materials. The results demonstrate that nuclear and electronic energy loss can lead to additive effects on irradiation damage production in some materials; while inmore » other materials, the competitive effects of electronic energy loss leads to recovery of damage induced by elastic collision cascades. Lastly, these results have significant implications for ion beam modification of materials, non-thermal recovery of ion implantation damage, and the response of materials to extreme radiation environments.« less

Properties of iron under core conditions

NASA Astrophysics Data System (ADS)

Brown, J. M.

2003-04-01

Underlying an understanding of the geodynamo and evolution of the core is knowledge of the physical and chemical properties of iron and iron mixtures under high pressure and temperature conditions. Key properties include the viscosity of the fluid outer core, thermal diffusivity, equations-of-state, elastic properties of solid phases, and phase equilibria for iron and iron-dominated mixtures. As is expected for work that continues to tax technological and intellectual limits, controversy has followed both experimental and theoretical progress in this field. However, estimates for the melting temperature of the inner core show convergence and the equation-of-state for iron as determined in independent experiments and theories are in remarkable accord. Furthermore, although the structure and elastic properties of the solid inner-core phase remains uncertain, theoretical and experimental underpinnings are better understood and substantial progress is likely in the near future. This talk will focus on an identification of properties that are reasonably well known and those that merit further detailed study. In particular, both theoretical and experimental (static and shock wave) determinations of the density of iron under extreme conditions are in agreement at the 1% or better level. The behavior of the Gruneisen parameter (which determines the geothermal gradient and controls much of the outer core heat flux) is constrained by experiment and theory under core conditions for both solid and liquid phases. Recent experiments and theory are suggestive of structure or structures other than the high-pressure hexagonal close-packed (HCP) phase. Various theories and experiments for the elasticity of HCP iron remain in poor accord. Uncontroversial constraints on core chemistry will likely never be possible. However, reasonable bounds are possible on the basis of seismic profiles, geochemical arguments, and determinations of sound velocities and densities at high pressure and temperature.

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turinsky, Paul J., E-mail: turinsky@ncsu.edu; Kothe, Douglas B., E-mail: kothe@ornl.gov

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear powermore » industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M&S capabilities, which is in progress, will assist in addressing long-standing and future operational and safety challenges of the nuclear industry. - Highlights: • Complexity of physics based modeling of light water reactor cores being addressed. • Capability developed to help address problems that have challenged the nuclear power industry. • Simulation capabilities that take advantage of high performance computing developed.« less

Nondestrucive analysis of fuel pins

DOEpatents

Stepan, I.E.; Allard, N.P.; Suter, C.R.

1972-11-03

Disclosure is made of a method and a correspondingly adapted facility for the nondestructive analysis of the concentation of fuel and poison in a nuclear reactor fuel pin. The concentrations of fuel and poison in successive sections along the entire length of the fuel pin are determined by measuring the reactivity of a thermal reactor as each successive small section of the fuel pin is exposed to the neutron flux of the reactor core and comparing the measured reactivity with the reactivities measured for standard fuel pins having various known concentrations. Only a small section of the length of the fuel pin is exposed to the neutron flux at any one time while the remainder of the fuel pin is shielded from the neutron flux. In order to expose only a small section at any one time, a boron-10-lined dry traverse tube is passed through the test region within the core of a low-power thermal nuclear reactor which has a very high fuel sensitivity. A narrow window in the boron-10 lining is positioned at the core center line. The fuel pins are then systematically traversed through the tube past the narrow window such that successive small sections along the length of the fuel pin are exposed to the neutron flux which passes through the narrow window.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Integrated NTP Vehicle Radiation Design

NASA Technical Reports Server (NTRS)

Caffrey, Jarvis A.; Rodriquez, Mitchell A.

2018-01-01

The development of a nuclear thermal propulsion stage requires consideration for radiation emitted from the nuclear reactor core. Applying shielding mass is an effective mitigating solution, but a better alternative is to incorporate some mitigation strategies into the propulsion stage and crew habitat. In this way, the required additional mass is minimized and the mass that must be applied may in some cases be able to serve multiple purposes. Strategies for crew compartment shielding are discussed that reduce dose from both engine and cosmic sources, and in some cases may also serve to reduce life support risks by permitting abundant water reserves. Early consideration for integrated mitigation solutions in a crewed nuclear thermal propulsion (NTP) vehicle will enable reduced radiation burden from both cosmic and nuclear sources, improved thrust-to-weight ratio or payload capacity by reducing 'dead mass' of shielding, and generally support a more robust risk posture for a NTP-powered Mars mission by permitting shorter trip times and increased water reserves.

Integrated NTP Vehicle Radiation Design

NASA Technical Reports Server (NTRS)

Caffrey, Jarvis; Rodriquez, Mitchell

2018-01-01

The development of a nuclear thermal propulsion stage requires consideration for radiation emitted from the nuclear reactor core. Applying shielding mass is an effective mitigating solution, but a better alternative is to incorporate some mitigation strategies into the propulsion stage and crew habitat. In this way, the required additional mass is minimized and the mass that must be applied may in some cases be able to serve multiple purposes. Strategies for crew compartment shielding are discussed that reduce dose from both engine and cosmic sources, and in some cases may also serve to reduce life support risks by permitting abundant water reserves. Early consideration for integrated mitigation solutions in a crewed nuclear thermal propulsion (NTP) vehicle will enable reduced radiation burden from both cosmic and nuclear sources, improved thrust-to-weight ratio or payload capacity by reducing 'dead mass' of shielding, and generally support a more robust risk posture for a NTP-powered Mars mission by permitting shorter trip times and increased water reserves

Satellite Gravity and the Geosphere: Contributions to the Study of the Solid Earth and Its Fluid Earth

NASA Technical Reports Server (NTRS)

Dickey, J. O.; Bentley, C. R.; Bilham, R.; Carton, J. A.; Eanes, R. J.; Herring, T. A.; Kaula, W. M.; Lagerloef, G. S. E.; Rojstaczer, S.; Smith, W. H. F.;

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

Energy Distribution of Electrons in Radiation Induced-Helium Plasmas. Ph.D. Thesis

NASA Technical Reports Server (NTRS)

Lo, R. H.

1972-01-01

Energy distribution of high energy electrons as they slow down and thermalize in a gaseous medium is studied. The energy distribution in the entire energy range from source energies down is studied analytically. A helium medium in which primary electrons are created by the passage of heavy-charged particles from nuclear reactions is emphasized. A radiation-induced plasma is of interest in a variety of applications, such as radiation pumped lasers and gaseous core nuclear reactors.

3D CFD ELECTROCHEMICAL AND HEAT TRANSFER MODEL OF AN INTERNALLY MANIFOLDED SOLID OXIDE ELECTROLYSIS CELL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grant L. Hawkes; James E. O'Brien; Greg Tao

2011-11-01

A three-dimensional computational fluid dynamics (CFD) electrochemical model has been created to model high-temperature electrolysis cell performance and steam electrolysis in an internally manifolded planar solid oxide electrolysis cell (SOEC) stack. This design is being evaluated at the Idaho National Laboratory for hydrogen production from nuclear power and process heat. Mass, momentum, energy, and species conservation and transport are provided via the core features of the commercial CFD code FLUENT. A solid-oxide fuel cell (SOFC) model adds the electrochemical reactions and loss mechanisms and computation of the electric field throughout the cell. The FLUENT SOFC user-defined subroutine was modified formore » this work to allow for operation in the SOEC mode. Model results provide detailed profiles of temperature, operating potential, steam-electrode gas composition, oxygen-electrode gas composition, current density and hydrogen production over a range of stack operating conditions. Single-cell and five-cell results will be presented. Flow distribution through both models is discussed. Flow enters from the bottom, distributes through the inlet plenum, flows across the cells, gathers in the outlet plenum and flows downward making an upside-down ''U'' shaped flow pattern. Flow and concentration variations exist downstream of the inlet holes. Predicted mean outlet hydrogen and steam concentrations vary linearly with current density, as expected. Effects of variations in operating temperature, gas flow rate, oxygen-electrode and steam-electrode current density, and contact resistance from the base case are presented. Contour plots of local electrolyte temperature, current density, and Nernst potential indicate the effects of heat transfer, reaction cooling/heating, and change in local gas composition. Results are discussed for using this design in the electrolysis mode. Discussion of thermal neutral voltage, enthalpy of reaction, hydrogen production, cell thermal efficiency, cell electrical efficiency, and Gibbs free energy are discussed and reported herein.« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anjos, J. C.; Barbosa, A. F.; Lima, H. P. Jr.

2010-03-30

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in amore » first step, to use the measured neutrino event rate to monitor the on--off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

NASA Astrophysics Data System (ADS)

Anjos, J. C.; Barbosa, A. F.; Bezerra, T. J. C.; Chimenti, P.; Gonzalez, L. F. G.; Kemp, E.; de Oliveira, M. A. Leigui; Lima, H. P.; Lima, R. M.; Nunokawa, H.

2010-03-01

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in a first step, to use the measured neutrino event rate to monitor the on—off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.

Nuclear reactor safety device

DOEpatents

Hutter, Ernest

1986-01-01

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

Space nuclear power systems 1989; Proceedings of the 6th Symposium, Albuquerque, NM, Jan. 8-12, 1989. Vols. 1 & 2

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1992-01-01

The present conference discusses such space nuclear power (SNP) issues as current design trends for SDI applications, ultrahigh heat-flux systems with curved surface subcooled nucleate boiling, design and manufacturing alternatives for low cost production of SNPs, a lightweight radioisotope heater for the Galileo mission, compatible materials for uranium fluoride-based gas core SNPs, Johnson noise thermometry for SNPs, and uranium nitride/rhenium compatibility studies for the SP-100 SNP. Also discussed are system issues in antimatter energy conversion, the thermal design of a heat source for a Brayton cycle radioisotope power system, structural and thermal analyses of an isotope heat source, a novel plant protection strategy for transient reactors, and beryllium toxicity.

A study of the required Rayleigh number to sustain dynamo with various inner core radius

NASA Astrophysics Data System (ADS)

Nishida, Y.; Katoh, Y.; Matsui, H.; Kumamoto, A.

2017-12-01

It is widely accepted that the geomagnetic field is sustained by thermal and compositional driven convections of a liquid iron alloy in the outer core. The generation process of the geomagnetic field has been studied by a number of MHD dynamo simulations. Recent studies of the ratio of the Earth's core evolution suggest that the inner solid core radius ri to the outer liquid core radius ro changed from ri/ro = 0 to 0.35 during the last one billion years. There are some studies of dynamo in the early Earth with smaller inner core than the present. Heimpel et al. (2005) revealed the Rayleigh number Ra of the onset of dynamo process as a function of ri/ro from simulation, while paleomagnetic observation shows that the geomagnetic field has been sustained for 3.5 billion years. While Heimpel and Evans (2013) studied dynamo processes taking into account the thermal history of the Earth's interior, there were few cases corresponding to the early Earth. Driscoll (2016) performed a series of dynamo based on a thermal evolution model. Despite a number of dynamo simulations, dynamo process occurring in the interior of the early Earth has not been fully understood because the magnetic Prandtl numbers in these simulations are much larger than that for the actual outer core.In the present study, we performed thermally driven dynamo simulations with different aspect ratio ri/ro = 0.15, 0.25 and 0.35 to evaluate the critical Ra for the thermal convection and required Ra to maintain the dynamo. For this purpose, we performed simulations with various Ra and fixed the other control parameters such as the Ekman, Prandtl, and magnetic Prandtl numbers. For the initial condition and boundary conditions, we followed the dynamo benchmark case 1 by Christensen et al. (2001). The results show that the critical Ra increases with the smaller aspect ratio ri/ro. It is confirmed that larger amplitude of buoyancy is required in the smaller inner core to maintain dynamo.

SASS-1--SUBASSEMBLY STRESS SURVEY CODE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Friedrich, C.M.

1960-01-01

SASS-1, an IBM-704 FORTRAN code, calculates pressure, thermal, and combined stresses in a nuclear reactor core subassembly. In addition to cross- section stresses, the code calculates axial shear stresses needed to keep plane cross sections plane under axial variations of temperature. The input and output nomenclature, arrangement, and formats are described. (B.O.G.)

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

EXPERIMENTAL METHODS TO ESTIMATE ACCUMULATED SOLIDS IN NUCLEAR WASTE TANKS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duignan, M.; Steeper, T.; Steimke, J.

2012-12-10

The Department of Energy has a large number of nuclear waste tanks. It is important to know if fissionable materials can concentrate when waste is transferred from staging tanks prior to feeding waste treatment plants. Specifically, there is a concern that large, dense particles, e.g., plutonium containing, could accumulate in poorly mixed regions of a blend tank heel for tanks that employ mixing jet pumps. At the request of the DOE Hanford Tank Operations Contractor, Washington River Protection Solutions, the Engineering Development Laboratory of the Savannah River National Laboratory performed a scouting study in a 1/22-scale model of a wastemore » tank to investigate this concern and to develop measurement techniques that could be applied in a more extensive study at a larger scale. Simulated waste tank solids and supernatant were charged to the test tank and rotating liquid jets were used to remove most of the solids. Then the volume and shape of the residual solids and the spatial concentration profiles for the surrogate for plutonium were measured. This paper discusses the overall test results, which indicated heavy solids only accumulate during the first few transfer cycles, along with the techniques and equipment designed and employed in the test. Those techniques include: Magnetic particle separator to remove stainless steel solids, the plutonium surrogate from a flowing stream; Magnetic wand used to manually remove stainless steel solids from samples and the tank heel; Photographs were used to determine the volume and shape of the solids mounds by developing a composite of topographical areas; Laser rangefinders to determine the volume and shape of the solids mounds; Core sampler to determine the stainless steel solids distribution within the solids mounds; Computer driven positioner that placed the laser rangefinders and the core sampler over solids mounds that accumulated on the bottom of a scaled staging tank in locations where jet velocities were low. These devices and techniques were very effective to estimate the movement, location, and concentrations of the solids representing plutonium and are expected to perform well at a larger scale. The operation of the techniques and their measurement accuracies will be discussed as well as the overall results of the accumulated solids test.« less

Experimental Methods to Estimate Accumulated Solids in Nuclear Waste Tanks - 13313

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duignan, Mark R.; Steeper, Timothy J.; Steimke, John L.

2013-07-01

The Department of Energy has a large number of nuclear waste tanks. It is important to know if fissionable materials can concentrate when waste is transferred from staging tanks prior to feeding waste treatment plants. Specifically, there is a concern that large, dense particles, e.g., plutonium containing, could accumulate in poorly mixed regions of a blend tank heel for tanks that employ mixing jet pumps. At the request of the DOE Hanford Tank Operations Contractor, Washington River Protection Solutions, the Engineering Development Laboratory of the Savannah River National Laboratory performed a scouting study in a 1/22-scale model of a wastemore » tank to investigate this concern and to develop measurement techniques that could be applied in a more extensive study at a larger scale. Simulated waste tank solids and supernatant were charged to the test tank and rotating liquid jets were used to remove most of the solids. Then the volume and shape of the residual solids and the spatial concentration profiles for the surrogate for plutonium were measured. This paper discusses the overall test results, which indicated heavy solids only accumulate during the first few transfer cycles, along with the techniques and equipment designed and employed in the test. Those techniques include: - Magnetic particle separator to remove stainless steel solids, the plutonium surrogate from a flowing stream. - Magnetic wand used to manually remove stainless steel solids from samples and the tank heel. - Photographs were used to determine the volume and shape of the solids mounds by developing a composite of topographical areas. - Laser range finders to determine the volume and shape of the solids mounds. - Core sampler to determine the stainless steel solids distribution within the solids mounds. - Computer driven positioner that placed the laser range finders and the core sampler over solids mounds that accumulated on the bottom of a scaled staging tank in locations where jet velocities were low. These devices and techniques were very effective to estimate the movement, location, and concentrations of the solids representing plutonium and are expected to perform well at a larger scale. The operation of the techniques and their measurement accuracies will be discussed as well as the overall results of the accumulated solids test. (authors)« less

Testing in Support of Fission Surface Power System Qualification

NASA Technical Reports Server (NTRS)

Houts, Mike; Bragg-Sitton, Shannon; Godfroy, Tom; Martin, Jim; Pearson, Boise; VanDyke, Melissa

2007-01-01

The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE reactors to ensure adequate radiation resistance. Facilities also exist for performing warm and cold zero-power criticals.

The WPI reactor-readying for the next generation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Morphology of the ferritin iron core by aberration corrected scanning transmission electron microscopy

NASA Astrophysics Data System (ADS)

Jian, Nan; Dowle, Miriam; Horniblow, Richard D.; Tselepis, Chris; Palmer, Richard E.

2016-11-01

As the major iron storage protein, ferritin stores and releases iron for maintaining the balance of iron in fauna, flora, and bacteria. We present an investigation of the morphology and iron loading of ferritin (from equine spleen) using aberration-corrected high angle annular dark field scanning transmission electron microscopy. Atom counting method, with size selected Au clusters as mass standards, was employed to determine the number of iron atoms in the nanoparticle core of each ferritin protein. Quantitative analysis shows that the nuclearity of iron atoms in the mineral core varies from a few hundred iron atoms to around 5000 atoms. Moreover, a relationship between the iron loading and iron core morphology is established, in which mineral core nucleates from a single nanoparticle, then grows along the protein shell before finally forming either a solid or hollow core structure.

Molten salts and nuclear energy production

NASA Astrophysics Data System (ADS)

Le Brun, Christian

2007-01-01

Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed.

Square lattice honeycomb reactor for space power and propulsion

NASA Astrophysics Data System (ADS)

Gouw, Reza; Anghaie, Samim

2000-01-01

The most recent nuclear design study at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) is the Moderated Square-Lattice Honeycomb (M-SLHC) reactor design utilizing the solid solution of ternary carbide fuels. The reactor is fueled with solid solution of 93% enriched (U,Zr,Nb)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. The M-SLHC design is based on a cylindrical core that has critical radius and length of 37 cm and 50 cm, respectively. This design utilized zirconium hydrate to act as moderator. The fuel sub-assemblies are designed as cylindrical tubes with 12 cm in diameter and 10 cm in length. Five fuel subassemblies are stacked up axially to form one complete fuel assembly. These fuel assemblies are then arranged in the circular arrangement to form two fuel regions. The first fuel region consists of six fuel assemblies, and 18 fuel assemblies for the second fuel region. A 10-cm radial beryllium reflector in addition to 10-cm top axial beryllium reflector is used to reduce neutron leakage from the system. To perform nuclear design analysis of the M-SLHC design, a series of neutron transport and diffusion codes are used. To optimize the system design, five axial regions are specified. In each axial region, temperature and fuel density are varied. The axial and radial power distributions for the system are calculated, as well as the axial and radial flux distributions. Temperature coefficients of the system are also calculated. A water submersion accident scenario is also analyzed for these systems. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel, which provides a relatively high thrust to weight ratio. .

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

NASA Technical Reports Server (NTRS)

Kirk, Daniel R.

2005-01-01

When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

Melt coaxial electrospinning: a versatile method for the encapsulation of solid materials and fabrication of phase change nanofibers.

PubMed

McCann, Jesse T; Marquez, Manuel; Xia, Younan

2006-12-01

We have developed a method based on melt coaxial electrospinning for fabricating phase change nanofibers consisting of long-chain hydrocarbon cores and composite sheaths. This method combines melt electrospinning with a coaxial spinneret and allows for nonpolar solids such as paraffins to be electrospun and encapsulated in one step. Shape-stabilized, phase change nanofibers have many potential applications as they are able to absorb, hold, and release large amounts of thermal energy over a certain temperature range by taking advantage of the large heat of fusion of long-chain hydrocarbons. We have focused on compounds with melting points near room temperature (octadecane) and body temperature (eicosane) as these temperature ranges are most valuable in practice. We have produced thermally stable, phase change materials up to 45 wt % octadecane, as measured by differential scanning calorimetry. In addition, the resultant fibers display novel segmented morphologies for the cores due to the rapid solidification of the hydrocarbons driven by evaporative cooling of the carrier solution. Aside from the fabrication of phase change nanofibers, the melt coaxial method is promising for applications related to microencapsulation and controlled release of drugs.

Nuclear reactor alignment plate configuration

DOEpatents

Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

2014-01-28

An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Gaseous fuel nuclear reactor research

NASA Technical Reports Server (NTRS)

Schwenk, F. C.; Thom, K.

1975-01-01

Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

Nuclear reactor shutdown control rod assembly

DOEpatents

Bilibin, Konstantin

1988-01-01

A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

75 FR 53985 - Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Unit 3; Temporary...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-02

... are authorized by law, will not present an undue risk to public health or safety, and are consistent... Public Health and Safety The underlying purpose of 10 CFR 50.46 is to establish acceptance criteria for... (LOCA) and non-LOCA criteria, mechanical design, thermal hydraulics, seismic, core physics, and...

Four-phonon scattering significantly reduces intrinsic thermal conductivity of solids

NASA Astrophysics Data System (ADS)

Feng, Tianli; Lindsay, Lucas; Ruan, Xiulin

2017-10-01

For decades, the three-phonon scattering process has been considered to govern thermal transport in solids, while the role of higher-order four-phonon scattering has been persistently unclear and so ignored. However, recent quantitative calculations of three-phonon scattering have often shown a significant overestimation of thermal conductivity as compared to experimental values. In this Rapid Communication we show that four-phonon scattering is generally important in solids and can remedy such discrepancies. For silicon and diamond, the predicted thermal conductivity is reduced by 30% at 1000 K after including four-phonon scattering, bringing predictions in excellent agreement with measurements. For the projected ultrahigh-thermal conductivity material, zinc-blende BAs, a competitor of diamond as a heat sink material, four-phonon scattering is found to be strikingly strong as three-phonon processes have an extremely limited phase space for scattering. The four-phonon scattering reduces the predicted thermal conductivity from 2200 to 1400 W/m K at room temperature. The reduction at 1000 K is 60%. We also find that optical phonon scattering rates are largely affected, being important in applications such as phonon bottlenecks in equilibrating electronic excitations. Recognizing that four-phonon scattering is expensive to calculate, in the end we provide some guidelines on how to quickly assess the significance of four-phonon scattering, based on energy surface anharmonicity and the scattering phase space. Our work clears the decades-long fundamental question of the significance of higher-order scattering, and points out ways to improve thermoelectrics, thermal barrier coatings, nuclear materials, and radiative heat transfer.

Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

NASA Astrophysics Data System (ADS)

Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

2007-01-01

The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 1013 to 1015 n/cm2. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 1015 to 1016 n/cm2 with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

NASA Technical Reports Server (NTRS)

Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

2007-01-01

The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 10(exp 13) to 10(exp 15) n per square centimeters. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 10(exp 15) to 10(exp 16) n per square centimeters with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David

2014-03-01

Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the coldmore » source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and affects mechanical properties of aluminum including density, neutron irradiation hardening, swelling, and loss of ductility. Because slightly greater quantities of silicon will be produced in the cold source moderator vessel for the LEU core, these effects will be slightly greater for the LEU core than for the HEU core. Three-group (thermal, epithermal, and fast) neutron flux results tallied in the cold source LH2 hemisphere show greater values for the LEU core under both BOC and EOC conditions. The thermal neutron flux in the LH2 hemisphere for the LEU core is about 12.4% greater at BOC and 2.7% greater at EOC than for the HEU core. Therefore, cold neutron scattering will not be adversely affected and the 4 12 neutrons conveyed to the cold neutron guide hall for research applications will be enhanced.« less

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Nuclear Analysis

NASA Technical Reports Server (NTRS)

Clement, J. D.; Kirby, K. D.

1973-01-01

Exploratory calculations were performed for several gas core breeder reactor configurations. The computational method involved the use of the MACH-1 one dimensional diffusion theory code and the THERMOS integral transport theory code for thermal cross sections. Computations were performed to analyze thermal breeder concepts and nonbreeder concepts. Analysis of breeders was restricted to the (U-233)-Th breeding cycle, and computations were performed to examine a range of parameters. These parameters include U-233 to hydrogen atom ratio in the gaseous cavity, carbon to thorium atom ratio in the breeding blanket, cavity size, and blanket size.

Atmospheric Mining in the Outer Solar System: Outer Planet In-Space Bases and Moon Bases for Resource Processing

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2017-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and deuterium can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and deuterium were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate, storage options, and different methods of direct use of the captured gases. While capturing 3He, large amounts of hydrogen and 4He are produced. The propulsion and transportation requirements for all of the major moons of Uranus and Neptune are presented. Analyses of orbital transfer vehicles (OTVs), landers, factories, and the issues with in-situ resource utilization (ISRU) low gravity processing factories are included. Preliminary observations are presented on near-optimal selections of moon base orbital locations, OTV power levels, and OTV and lander rendezvous points. Several artificial gravity in-space base designs and orbital sites at Uranus and Neptune and the OTV requirements to support them are also addressed.

Atmospheric Mining in the Outer Solar System: Resource Capturing, Exploration, and Exploitation

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2015-01-01

Atmospheric mining in the outer solar system (AMOSS) has been investigated as a means of fuel production for high-energy propulsion and power. Fusion fuels such as helium 3 (He-3) and hydrogen can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. 3He and hydrogen (deuterium, etc.) were the primary gases of interest, with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of AMOSS. These analyses included the gas capturing rate, storage options, and different methods of direct use of the captured gases. Additional supporting analyses were conducted to illuminate vehicle sizing and orbital transportation issues. While capturing 3He, large amounts of hydrogen and helium 4 (He-4) are produced. With these two additional gases, the potential exists for fueling small and large fleets of additional exploration and exploitation vehicles. Additional aerospacecraft or other aerial vehicles (UAVs, balloons, rockets, etc.) could fly through the outer-planet atmosphere to investigate cloud formation dynamics, global weather, localized storms or other disturbances, wind speeds, the poles, and so forth. Deep-diving aircraft (built with the strength to withstand many atmospheres of pressure) powered by the excess hydrogen or 4He may be designed to probe the higher density regions of the gas giants.

Solution and Solid State Nuclear Magnetic Resonance Spectroscopic Characterization of Efavirenz.

PubMed

Sousa, Eduardo Gomes Rodrigues de; Carvalho, Erika Martins de; San Gil, Rosane Aguiar da Silva; Santos, Tereza Cristina Dos; Borré, Leandro Bandeira; Santos-Filho, Osvaldo Andrade; Ellena, Javier

2016-09-01

Samples of efavirenz (EFZ) were evaluated to investigate the influence of the micronization process on EFZ stability. A combination of X-ray diffraction, thermal analysis, FTIR, observations of isotropic chemical shifts of (1)H in distinct solvents, their temperature dependence and spin-lattice relaxation time constants (T1), solution (1D and 2D) (13)C nuclear magnetic resonance (NMR), and solid-state (13)C NMR (CPMAS NMR) provides valuable structural information and structural elucidation of micronized EFZ and heptane-recrystallized polymorphs (EFZ/HEPT). This study revealed that the micronization process did not affect the EFZ crystalline structure. It was observed that the structure of EFZ/HEPT is in the same form as that obtained from ethyl acetate/hexane, as shown in the literature. A comparison of the solid-state NMR spectra revealed discrepancies regarding the assignments of some carbons published in the literature that have been resolved. Copyright © 2016 American Pharmacists Association®. Published by Elsevier Inc. All rights reserved.

Nuclear fuel in a reactor accident.

PubMed

Burns, Peter C; Ewing, Rodney C; Navrotsky, Alexandra

2012-03-09

Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Nuclear Power Plant environment`s surveillance by satellite remote sensing and in-situ monitoring data

NASA Astrophysics Data System (ADS)

Zoran, Maria

The main environmental issues affecting the broad acceptability of nuclear power plant are the emission of radioactive materials, the generation of radioactive waste, and the potential for nuclear accidents. All nuclear fission reactors, regardless of design, location, operator or regulator, have the potential to undergo catastrophic accidents involving loss of control of the reactor core, failure of safety systems and subsequent widespread fallout of hazardous fission products. Risk is the mathematical product of probability and consequences, so lowprobability and high-consequence accidents, by definition, have a high risk. NPP environment surveillance is a very important task in frame of risk assessment. Satellite remote sensing data had been applied for dosimeter levels first time for Chernobyl NPP accident in 1986. Just for a normal functioning of a nuclear power plant, multitemporal and multispectral satellite data in complementarily with field data are very useful tools for NPP environment surveillance and risk assessment. Satellite remote sensing is used as an important technology to help environmental research to support research analysis of spatio-temporal dynamics of environmental features nearby nuclear facilities. Digital processing techniques applied to several LANDSAT, MODIS and QuickBird data in synergy with in-situ data are used to assess the extent and magnitude of radiation and non-radiation effects on the water, near field soil, vegetation and air. As a test case the methodology was applied for for Nuclear Power Plant (NPP) Cernavoda, Romania. Thermal discharge from nuclear reactors cooling is dissipated as waste heat in Danube-Black -Sea Canal and Danube River. Water temperatures captured in thermal IR imagery are correlated with meteorological parameters. If during the winter thermal plume is localized to an area of a few km of NPP, the temperature difference between the plume and non-plume areas being about 1.5 oC, during summer and fall , is a larger thermal plume up to 5-6 km far along Danube Black Sea Canal ,the temperature change is about 1.0 oC.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolstad, J.W.; Haarman, R.A.

The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid. Calculations show that continued injection of high pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island.

Computer Simulation To Assess The Feasibility Of Coring Magma

NASA Astrophysics Data System (ADS)

Su, J.; Eichelberger, J. C.

2017-12-01

Lava lakes on Kilauea Volcano, Hawaii have been successfully cored many times, often with nearly complete recovery and at temperatures exceeding 1100oC. Water exiting nozzles on the diamond core bit face quenches melt to glass just ahead of the advancing bit. The bit readily cuts a clean annulus and the core, fully quenched lava, passes smoothly into the core barrel. The core remains intact after recovery, even when there are comparable amounts of glass and crystals with different coefficients of thermal expansion. The unique resulting data reveal the rate and sequence of crystal growth in cooling basaltic lava and the continuous liquid line of descent as a function of temperature from basalt to rhyolite. Now that magma bodies, rather than lava pooled at the surface, have been penetrated by geothermal drilling, the question arises as to whether similar coring could be conducted at depth, providing fundamentally new insights into behavior of magma. This situation is considerably more complex because the coring would be conducted at depths exceeding 2 km and drilling fluid pressures of 20 MPa or more. Criteria that must be satisfied include: 1) melt is quenched ahead of the bit and the core itself must be quenched before it enters the barrel; 2) circulating drilling fluid must keep the temperature of the coring assembling cooled to within operational limits; 3) the drilling fluid column must nowhere exceed the local boiling point. A fluid flow simulation was conducted to estimate the process parameters necessary to maintain workable temperatures during the coring operation. SolidWorks Flow Simulation was used to estimate the effect of process parameters on the temperature distribution of the magma immediately surrounding the borehole and of drilling fluid within the bottom-hole assembly (BHA). A solid model of the BHA was created in SolidWorks to capture the flow behavior around the BHA components. Process parameters used in the model include the fluid properties and temperature of magma, coolant flow rate, rotation speed, and rate of penetration (ROP). The modeling results indicate that there are combinations of process parameters that will provide sufficient cooling to enable the desired coring process in magma.

Powder Processing of High Temperature Cermets and Carbides at Marshall Space Flight Center

NASA Technical Reports Server (NTRS)

Salvail, Pat; Panda, Binayak; Hickman, Robert R.

2007-01-01

The Materials and Processing Laboratory at NASA Marshall Space Flight Center is developing Powder Metallurgy (PM) processing techniques for high temperature cermet and carbide material consolidation. These new group of materials would be utilized in the nuclear core for Nuclear Thermal Rockets (NTR). Cermet materials offer several advantages for NTR such as retention of fission products and fuels, better thermal shock resistance, hydrogen compatibility, high thermal conductivity, and high strength. Carbide materials offer the highest operating temperatures but are sensitive to thermal stresses and are difficult to process. To support the effort, a new facility has been setup to process refractory metal, ceramic, carbides and depleted uranium-based powders. The facility inciudes inert atmosphere glove boxes for the handling of reactive powders, a high temperature furnace, and powder processing equipment used for blending, milling, and sieving. The effort is focused on basic research to identify the most promising compositions and processing techniques. Several PM processing methods including Cold and Hot Isostatic Pressing are being evaluated to fabricate samples for characterization and hot hydrogen testing.

Resonance-inclined optical nuclear spin polarization of liquids in diamond structures

NASA Astrophysics Data System (ADS)

Chen, Q.; Schwarz, I.; Jelezko, F.; Retzker, A.; Plenio, M. B.

2016-02-01

Dynamic nuclear polarization (DNP) of molecules in a solution at room temperature has the potential to revolutionize nuclear magnetic resonance spectroscopy and imaging. The prevalent methods for achieving DNP in solutions are typically most effective in the regime of small interaction correlation times between the electron and nuclear spins, limiting the size of accessible molecules. To solve this limitation, we design a mechanism for DNP in the liquid phase that is applicable for large interaction correlation times. Importantly, while this mechanism makes use of a resonance condition similar to solid-state DNP, the polarization transfer is robust to a relatively large detuning from the resonance due to molecular motion. We combine this scheme with optically polarized nitrogen-vacancy (NV) center spins in nanodiamonds to design a setup that employs optical pumping and is therefore not limited by room temperature electron thermal polarization. We illustrate numerically the effectiveness of the model in a flow cell containing nanodiamonds immobilized in a hydrogel, polarizing flowing water molecules 4700-fold above thermal polarization in a magnetic field of 0.35 T, in volumes detectable by current NMR scanners.

Quantum phases of dipolar soft-core bosons

NASA Astrophysics Data System (ADS)

Grimmer, D.; Safavi-Naini, A.; Capogrosso-Sansone, B.; Söyler, Ş. G.

2014-10-01

We study the phase diagram of a system of soft-core dipolar bosons confined to a two-dimensional optical lattice layer. We assume that dipoles are aligned perpendicular to the layer such that the dipolar interactions are purely repulsive and isotropic. We consider the full dipolar interaction and perform path-integral quantum Monte Carlo simulations using the worm algorithm. Besides a superfluid phase, we find various solid and supersolid phases. We show that, unlike what was found previously for the case of nearest-neighbor interaction, supersolid phases are stabilized by doping the solids not only with particles but with holes as well. We further study the stability of these quantum phases against thermal fluctuations. Finally, we discuss pair formation and the stability of the pair checkerboard phase formed in a bilayer geometry, and we suggest experimental conditions under which the pair checkerboard phase can be observed.

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

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.

Nootkatone encapsulation by cyclodextrins: Effect on water solubility and photostability.

PubMed

Kfoury, Miriana; Landy, David; Ruellan, Steven; Auezova, Lizette; Greige-Gerges, Hélène; Fourmentin, Sophie

2017-12-01

Nootkatone (NO) is a sesquiterpenoid volatile flavor, used in foods, cosmetics and pharmaceuticals, possessing also insect repellent activity. Its application is limited because of its low aqueous solubility and stability; this could be resolved by encapsulation in cyclodextrins (CDs). This study evaluated the encapsulation of NO by CDs using phase solubility studies, Isothermal Titration Calorimetry, Nuclear Magnetic Resonance spectroscopy and molecular modeling. Solid CD/NO inclusion complex was prepared and characterized for encapsulation efficiency and loading capacity using UV-Visible. Thermal properties were investigated by thermogravimetric-differential thermal analysis and release studies were performed using multiple headspace extraction. Formation constants (K f ) proved the formation of stable inclusion complexes. NO aqueous solubility, photo- and thermal stability were enhanced and the release could be insured from solid complex in aqueous solution. This suggests that CDs are promising carrier to improve NO properties and, consequently, to enlarge its use in foods, cosmetics, pharmaceuticals and agrochemicals preparations. Copyright © 2016 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haugen, Carl C.; Forget, Benoit; Smith, Kord S.

Most high performance computing systems being deployed currently and envisioned for the future are based on making use of heavy parallelism across many computational nodes and many concurrent cores. These types of heavily parallel systems often have relatively little memory per core but large amounts of computing capability. This places a significant constraint on how data storage is handled in many Monte Carlo codes. This is made even more significant in fully coupled multiphysics simulations, which requires simulations of many physical phenomena be carried out concurrently on individual processing nodes, which further reduces the amount of memory available for storagemore » of Monte Carlo data. As such, there has been a move towards on-the-fly nuclear data generation to reduce memory requirements associated with interpolation between pre-generated large nuclear data tables for a selection of system temperatures. Methods have been previously developed and implemented in MIT’s OpenMC Monte Carlo code for both the resolved resonance regime and the unresolved resonance regime, but are currently absent for the thermal energy regime. While there are many components involved in generating a thermal neutron scattering cross section on-the-fly, this work will focus on a proposed method for determining the energy and direction of a neutron after a thermal incoherent inelastic scattering event. This work proposes a rejection sampling based method using the thermal scattering kernel to determine the correct outgoing energy and angle. The goal of this project is to be able to treat the full S (a, ß) kernel for graphite, to assist in high fidelity simulations of the TREAT reactor at Idaho National Laboratory. The method is, however, sufficiently general to be applicable in other thermal scattering materials, and can be initially validated with the continuous analytic free gas model.« less

Affordable Development and Demonstration of a Small Nuclear Thermal Rocket (NTR) Engine and Stage: How Small Is Big Enough?

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Sefcik, Robert J.; Fittje, James E.; McCurdy, David R.; Qualls, Arthur L.; Schnitzler, Bruce G.; Werner, James E.; Weitzberg, Abraham; Joyner, Claude R.

2016-01-01

The Nuclear Thermal Rocket (NTR) derives its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core. It generates high thrust and has a specific impulse potential of approximately 900 specific impulse - a 100 percent increase over today's best chemical rockets. The Nuclear Thermal Propulsion (NTP) project, funded by NASA's Advanced Exploration Systems (AES) program, includes five key task activities: (1) Recapture, demonstration, and validation of heritage graphite composite (GC) fuel (selected as the Lead Fuel option); (2) Engine Conceptual Design; (3) Operating Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable Development Strategy. During fiscal year (FY) 2014, a preliminary Design Development Test and Evaluation (DDT&E) plan and schedule for NTP development was outlined by the NASA Glenn Research Center (GRC), Department of Energy (DOE) and industry that involved significant system-level demonstration projects that included Ground Technology Demonstration (GTD) tests at the Nevada National Security Site (NNSS), followed by a Flight Technology Demonstration (FTD) mission. To reduce cost for the GTD tests and FTD mission, small NTR engines, in either the 7.5 or 16.5 kilopound-force thrust class, were considered. Both engine options used GC fuel and a common fuel element (FE) design. The small approximately 7.5 kilopound-force criticality-limited engine produces approximately157 thermal megawatts and its core is configured with parallel rows of hexagonal-shaped FEs and tie tubes (TTs) with a FE to TT ratio of approximately 1:1. The larger approximately 16.5 kilopound-force Small Nuclear Rocket Engine (SNRE), developed by Los Alamos National Laboratory (LANL) at the end of the Rover program, produces approximately 367 thermal megawatts and has a FE to TT ratio of approximately 2:1. Although both engines use a common 35-inch (approximately 89-centimeters) -long FE, the SNRE's larger diameter core contains approximately 300 more FEs needed to produce an additional 210 thermal megawatts of power. To reduce the cost of the FTD mission, a simple one-burn lunar flyby mission was considered to reduce the liquid hydrogen (LH2) propellant loading, the stage size and complexity. Use of existing and flight proven liquid rocket and stage hardware (e.g., from the RL10B-2 engine and Delta Cryogenic Second Stage) was also maximized to further aid affordability. This paper examines the pros and cons of using these two small engine options, including their potential to support future human exploration missions to the Moon, near Earth asteroids (NEA), and Mars, and recommends a preferred size. It also provides a preliminary assessment of the key activities, development options, and schedule required to affordably build, ground test and fly a small NTR engine and stage within a 10-year timeframe.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sherman, Andrew J

A heterogeneous body having ceramic rich cermet regions in a more ductile metal matrix. The heterogeneous bodies are formed by thermal spray operations on metal substrates. The thermal spray operations apply heat to a cermet powder and project it onto a solid substrate. The cermet powder is composed of complex composite particles in which a complex ceramic-metallic core particle is coated with a matrix precursor. The cermet regions are generally comprised of complex ceramic-metallic composites that correspond approximately to the core particles. The cermet regions are approximately lenticular shaped with an average width that is at least approximately twice themore » average thickness. The cermet regions are imbedded within the matrix phase and generally isolated from one another. They have obverse and reverse surfaces. The matrix phase is formed from the matrix precursor coating on the core particles. The amount of heat applied during the formation of the heterogeneous body is controlled so that the core particles soften but do not become so fluid that they disperse throughout the matrix phase. The force of the impact on the surface of the substrate tends to flatten them. The flattened cermet regions tend to be approximately aligned with one another in the body.« less

Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruce G. Schnitzler; Stanley K. Borowski

2010-07-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effortmore » was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.« less

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

The HALNA project: Diode-pumped solid-state laser for inertial fusion energy

NASA Astrophysics Data System (ADS)

Kawashima, T.; Ikegawa, T.; Kawanaka, J.; Miyanaga, N.; Nakatsuka, M.; Izawa, Y.; Matsumoto, O.; Yasuhara, R.; Kurita, T.; Sekine, T.; Miyamoto, M.; Kan, H.; Furukawa, H.; Motokoshi, S.; Kanabe, T.

2006-06-01

High-enery, rep.-rated, diode-pumped solid-state laser (DPSSL) is one of leading candidates for inertial fusion energy driver (IFE) and related laser-driven high-field applications. The project for the development of IFE laser driver in Japan, HALNA (High Average-power Laser for Nuclear Fusion Application) at ILE, Osaka University, aims to demonstrate 100-J pulse energy at 10 Hz rep. rate with 5 times diffraction limited beam quality. In this article, the advanced solid-state laser technologies for one half scale of HALNA (50 J, 10 Hz) are presented including thermally managed slab amplifier of Nd:phosphate glass and zig-zag optical geometry, and uniform, large-area diode-pumping.

Perturbation of nuclear spin polarizations in solid state NMR of nitroxide-doped samples by magic-angle spinning without microwaves.

PubMed

Thurber, Kent R; Tycko, Robert

2014-05-14

We report solid state (13)C and (1)H nuclear magnetic resonance (NMR) experiments with magic-angle spinning (MAS) on frozen solutions containing nitroxide-based paramagnetic dopants that indicate significant perturbations of nuclear spin polarizations without microwave irradiation. At temperatures near 25 K, (1)H and cross-polarized (13)C NMR signals from (15)N,(13)C-labeled L-alanine in trinitroxide-doped glycerol/water are reduced by factors as large as six compared to signals from samples without nitroxide doping. Without MAS or at temperatures near 100 K, differences between signals with and without nitroxide doping are much smaller. We attribute most of the reduction of NMR signals under MAS near 25 K to nuclear spin depolarization through the cross-effect dynamic nuclear polarization mechanism, in which three-spin flips drive nuclear polarizations toward equilibrium with spin polarization differences between electron pairs. When T1e is sufficiently long relative to the MAS rotation period, the distribution of electron spin polarization across the nitroxide electron paramagnetic resonance lineshape can be very different from the corresponding distribution in a static sample at thermal equilibrium, leading to the observed effects. We describe three-spin and 3000-spin calculations that qualitatively reproduce the experimental observations.

Early Stage of Origin of Earth (interval after Emergence of Sun, Formation of Liquid Core, Formation of Solid Core)

NASA Astrophysics Data System (ADS)

Pechernikova, G. V.; Sergeev, V. N.

2017-05-01

Gravitational collapse of interstellar molecular cloud fragment has led to the formation of the Sun and its surrounding protoplanetary disk, consisting of 5 × 10^5 dust and gas. The collapse continued (1 years. Age of solar system (about 4.57×10^9 years) determine by age calcium-aluminum inclusions (CAI) which are present at samples of some meteorites (chondrites). Subsidence of dust to the central plane of a protoplanetary disk has led to formation of a dust subdisk which as a result of gravitational instability has broken up to condensations. In the process of collisional evolution they turned into dense planetesimals from which the planets formed. The accounting of a role of large bodies in evolution of a protoplanetary swarm in the field of terrestrial planets has allowed to define times of formation of the massive bodies permitting their early differentiation at the expense of short-lived isotopes heating and impacts to the melting temperature of the depths. The total time of Earth's growth is estimated about 10^8 years. Hf geochronometer showed that the core of the Earth has existed for Using W about 3×10^7 Hf geohronometer years since the formation of the CAI. Thus data W point to the formation of the Earth's core during its accretion. The paleomagnetic data indicate the existence of Earth's magnetic field past 3.5×10^9 years. But the age of the solid core, estimated by heat flow at the core-mantle boundary is 1.7×10^9 (0.5 years). Measurements of the thermal conductivity of liquid iron under the conditions that exist in the Earth's core, indicate the absence of the need for a solid core of existence to support the work geodynamo, although electrical resistivity measurements yield the opposite result.

76 FR 26771 - Florida Power & Light Company; Turkey Point, Units 3 and 4; Notice of Consideration of Issuance...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-09

... Company; Turkey Point, Units 3 and 4; Notice of Consideration of Issuance of Amendment to Facility... issued to Florida Power & Light Co. (the licensee) for operation of the Turkey Point Nuclear Generating... licensed core power level for Turkey Point, Units 3 and 4, from 2300 megawatts thermal (MWt) to 2644 MWt...

Temperature Profile in Fuel and Tie-Tubes for Nuclear Thermal Propulsion Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vishal Patel

A finite element method to calculate temperature profiles in heterogeneous geometries of tie-tube moderated LEU nuclear thermal propulsion systems and HEU designs with tie-tubes is developed and implemented in MATLAB. This new method is compared to previous methods to demonstrate shortcomings in those methods. Typical methods to analyze peak fuel centerline temperature in hexagonal geometries rely on spatial homogenization to derive an analytical expression. These methods are not applicable to cores with tie-tube elements because conduction to tie-tubes cannot be accurately modeled with the homogenized models. The fuel centerline temperature directly impacts safety and performance so it must be predictedmore » carefully. The temperature profile in tie-tubes is also important when high temperatures are expected in the fuel because conduction to the tie-tubes may cause melting in tie-tubes, which may set maximum allowable performance. Estimations of maximum tie-tube temperature can be found from equivalent tube methods, however this method tends to be approximate and overly conservative. A finite element model of heat conduction on a unit cell can model spatial dependence and non-linear conductivity for fuel and tie-tube systems allowing for higher design fidelity of Nuclear Thermal Propulsion.« less

Temperature actuated shutdown assembly for a nuclear reactor

DOEpatents

Sowa, Edmund S.

1976-01-01

Three identical bimetallic disks, each shaped as a spherical cap with its convex side composed of a layer of metal such as molybdenum and its concave side composed of a metal of a relatively higher coefficient of thermal expansion such as stainless steel, are retained within flanges attached to three sides of an inner hexagonal tube containing a neutron absorber to be inserted into a nuclear reactor core. Each disk holds a metal ball against its normally convex side so that the ball projects partially through a hole in the tube located concentrically with the center of each disk; at a predetermined temperature an imbalance of thermally induced stresses in at least one of the disks will cause its convex side to become concave and its concave side to become convex, thus pulling the ball from the hole in which it is located. The absorber has a conical bottom supported by the three balls and is small enough in relation to the internal dimensions of the tube to allow it to slip toward the removed ball or balls, thus clearing the unremoved balls or ball so that it will fall into the reactor core.

Physical particularities of nuclear reactors using heavy moderators of neutrons

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.

2016-12-01

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using 233U as a fissile nuclide and 232Th and 231Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Elastically driven intermittent microscopic dynamics in soft solids

NASA Astrophysics Data System (ADS)

Bouzid, Mehdi; Colombo, Jader; Barbosa, Lucas Vieira; Del Gado, Emanuela

2017-06-01

Soft solids with tunable mechanical response are at the core of new material technologies, but a crucial limit for applications is their progressive aging over time, which dramatically affects their functionalities. The generally accepted paradigm is that such aging is gradual and its origin is in slower than exponential microscopic dynamics, akin to the ones in supercooled liquids or glasses. Nevertheless, time- and space-resolved measurements have provided contrasting evidence: dynamics faster than exponential, intermittency and abrupt structural changes. Here we use 3D computer simulations of a microscopic model to reveal that the timescales governing stress relaxation, respectively, through thermal fluctuations and elastic recovery are key for the aging dynamics. When thermal fluctuations are too weak, stress heterogeneities frozen-in upon solidification can still partially relax through elastically driven fluctuations. Such fluctuations are intermittent, because of strong correlations that persist over the timescale of experiments or simulations, leading to faster than exponential dynamics.

Enabling thermal processing of ritonavir-polyvinyl alcohol amorphous solid dispersions by KinetiSol® Dispersing.

PubMed

LaFountaine, Justin S; Jermain, Scott V; Prasad, Leena Kumari; Brough, Chris; Miller, Dave A; Lubda, Dieter; McGinity, James W; Williams, Robert O

2016-04-01

Polyvinyl alcohol has received little attention as a matrix polymer in amorphous solid dispersions (ASDs) due to its thermal and rheological limitations in extrusion processing and limited organic solubility in spray drying applications. Additionally, in extrusion processing, the high temperatures required to process often exclude thermally labile APIs. The purpose of this study was to evaluate the feasibility of processing polyvinyl alcohol amorphous solid dispersions utilizing the model compound ritonavir with KinetiSol® Dispersing (KSD) technology. The effects of KSD rotor speed and ejection temperature on the physicochemical properties of the processed material were evaluated. Powder X-ray diffraction and modulated differential scanning calorimetry were used to confirm amorphous conversion. Liquid chromatography-mass spectroscopy was used to characterize and identify degradation pathways of ritonavir during KSD processing and (13)C nuclear magnetic resonance spectroscopy was used to investigate polymer stability. An optimal range of processing conditions was found that resulted in amorphous product and minimal to no drug and polymer degradation. Drug release of the ASD produced from the optimal processing conditions was evaluated using a non-sink, pH-shift dissolution test. The ability to process amorphous solid dispersions with polyvinyl alcohol as a matrix polymer will enable further investigations of the polymer's performance in amorphous systems for poorly water-soluble compounds. Copyright © 2016 Elsevier B.V. All rights reserved.

Porous materials produced from incineration ash using thermal plasma technology.

PubMed

Yang, Sheng-Fu; Chiu, Wen-Tung; Wang, To-Mai; Chen, Ching-Ting; Tzeng, Chin-Ching

2014-06-01

This study presents a novel thermal plasma melting technique for neutralizing and recycling municipal solid waste incinerator (MSWI) ash residues. MSWI ash residues were converted into water-quenched vitrified slag using plasma vitrification, which is environmentally benign. Slag is adopted as a raw material in producing porous materials for architectural and decorative applications, eliminating the problem of its disposal. Porous materials are produced using water-quenched vitrified slag with Portland cement and foaming agent. The true density, bulk density, porosity and water absorption ratio of the foamed specimens are studied here by varying the size of the slag particles, the water-to-solid ratio, and the ratio of the weights of the core materials, including the water-quenched vitrified slag and cement. The thermal conductivity and flexural strength of porous panels are also determined. The experimental results show the bulk density and the porosity of the porous materials are 0.9-1.2 g cm(-3) and 50-60%, respectively, and the pore structure has a closed form. The thermal conductivity of the porous material is 0.1946 W m(-1) K(-1). Therefore, the slag composite materials are lightweight and thermal insulators having considerable potential for building applications. Copyright © 2013 Elsevier Ltd. All rights reserved.

Room-temperature in situ nuclear spin hyperpolarization from optically pumped nitrogen vacancy centres in diamond

DOE PAGES

King, Jonathan P.; Jeong, Keunhong; Vassiliou, Christophoros C.; ...

2015-12-07

Low detection sensitivity stemming from the weak polarization of nuclear spins is a primary limitation of magnetic resonance spectroscopy and imaging. Methods have been developed to enhance nuclear spin polarization but they typically require high magnetic fields, cryogenic temperatures or sample transfer between magnets. Here we report bulk, room-temperature hyperpolarization of 13C nuclear spins observed via high-field magnetic resonance. The technique harnesses the high optically induced spin polarization of diamond nitrogen vacancy centres at room temperature in combination with dynamic nuclear polarization. We observe bulk nuclear spin polarization of 6%, an enhancement of ~170,000 over thermal equilibrium. The signal ofmore » the hyperpolarized spins was detected in situ with a standard nuclear magnetic resonance probe without the need for sample shuttling or precise crystal orientation. In conclusion, hyperpolarization via optical pumping/dynamic nuclear polarization should function at arbitrary magnetic fields enabling orders of magnitude sensitivity enhancement for nuclear magnetic resonance of solids and liquids under ambient conditions.« less

Thermal evolution of a differentiated Ganymede and implications for surface features

NASA Technical Reports Server (NTRS)

Kirk, R. L.; Stevenson, D. J.

1987-01-01

Thermodynamic models are developed for the processes which controlled the evolution of the surface Ganymede, an icy Jovian satellite assumed to have a rock-rich core surrounded by a water-ice mantle. Account is taken of a heat pulse which would have arisen from a Rayleigh-Taylor instability at a deep-seated liquid-solid water interface, rapid fracturing from global stresses imposed by warm ice diapiric upwelling, impacts by large meteorites, and resurfacing by ice flows (rather than core formation). Comparisons are made with existing models for the evolution of Callisto, and the difficulties in defining a mechanism which produced the groove terrain of Ganymede are discussed.

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robertson, Sean; Dewan, Leslie; Massie, Mark

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parametersmore » necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.« less

Homogeneous synthesis of cellulose acrylate-g-poly (n-alkyl acrylate) solid-solid phase change materials via free radical polymerization.

PubMed

Qian, Yong-Qiang; Han, Na; Bo, Yi-Wen; Tan, Lin-Li; Zhang, Long-Fei; Zhang, Xing-Xiang

2018-08-01

A novel solid-solid phase change materials, namely, cellulose acrylate-g-poly (n-alkyl acrylate) (CA-g-PAn) (n = 14, 16 and 18) were successfully synthesized by free radical polymerization in N, N-dimethylacetamide (DMAc). The successful grafting was confirmed by fourier transform infrared spectra (FT-IR) and nuclear magnetic resonance (NMR). The properties of the CA-g-PAn copolymers were investigated by differential scanning calorimetry (DSC), thermogravimetric analysis (TGA). The phase change temperatures and the melting enthalpies of CA-g-PAn copolymers are in the range of 10.1-53.2 °C and 15-95 J/g, respectively. It can be adjusted by the contents of poly (n-alkyl acrylate) and the length of alkyl side-chain. The thermal resistant temperatures of CA-g-PA14, 16 and 18 copolymers are 308 °C, 292 °C and 273 °C, respectively. It show that all of grafting materials exhibit good thermal stability and shape stability. Therefore, it is expected to be applied in the cellulose-based thermos-regulating field. Copyright © 2018 Elsevier Ltd. All rights reserved.

Floquet Topological Insulators in Uranium Compounds

NASA Astrophysics Data System (ADS)

Pi, Shu-Ting; Savrasov, Sergey

2014-03-01

A major issue regarding the Uranium based nuclear fuels is to conduct the heat from the core area to its outer area. Unfortunately, those materials are notorious for their extremely low thermal conductivity due to the phonon-dominated-heat-transport properties in insulating states. Although metallic Uranium compounds are helpful in increasing the thermal conductivity, their low melting point still make those efforts in vain. In this report, we will figure out potential Uranium based Floquet topological insulators where the insulating bulk states accompanied with metallic surface states is achieved by applying periodic electrical fields which makes the coexistence of both benefits possible.

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

NASA Astrophysics Data System (ADS)

Turinsky, Paul J.; Kothe, Douglas B.

2016-05-01

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics ;core simulator; based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M&S capabilities, which is in progress, will assist in addressing long-standing and future operational and safety challenges of the nuclear industry.

PL-3, PHASE I, TASK 3, RESEARCH AND DEVELOPMENT REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humphries, G. E.

1962-03-12

Results of researeh and development tasks are presented along with recommendations for future development work Work (s reported ofn the areas of plant assembly and relocation, housings and footings, waste heat dissipation, instrumentation, refueling systems, waste disposal, shiceding, core nuclear thermal and hydraulic studies, gaseous waste processing, and critical experiments on a 5 x 5 array of Type 3 fuel elements. (auth)

Nuclear Power - Post Fukushima

NASA Astrophysics Data System (ADS)

Reyes, Jose, Jr.

2011-10-01

The extreme events that led to the prolonged power outage at the Fukushima Daiicchi nuclear plant have highlighted the importance of assuring a means for stable long term cooling of the nuclear fuel and containment following a complete station blackout. Legislative bodies, regulatory agencies and industry are drawing lessons from those events and considering what changes, if any, are needed to nuclear power, post Fukushima. The enhanced safety of a new class of reactor designed by NuScale Power is drawing significant attention in light of the Fukushima events. During normal operation, each NuScale containment is fully immersed in a water-filled stainless steel lined concrete pool that resides underground. The pool, housed in a Seismic Category I building, is large enough to provided 30 days of core and containment cooling without adding water. After 30 days, the decay heat generations coupled with thermal radiation heat transfer is completely adequate to remove core decay heat for an unlimited period of time. These passive power systems can perform their function without requiring an external supply of water of power. An assessment of the NuScale passive systems is being performed through a comprehensive test program that includes the NuScale integral system test facility at Oregon State University

Neutron dose estimation in a zero power nuclear reactor

NASA Astrophysics Data System (ADS)

Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

2016-10-01

This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

NASA Astrophysics Data System (ADS)

Ramdhani, Rahmi N.; Prastyo, Puguh A.; Waris, Abdul; Widayani; Kurniadi, Rizal

2017-07-01

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

The Thermal State of KS 1731-260 after 14.5 years in Quiescence

NASA Astrophysics Data System (ADS)

Merritt, Rachael L.; Cackett, Edward M.; Brown, Edward F.; Page, Dany; Cumming, Andrew; Degenaar, Nathalie; Deibel, Alex; Homan, Jeroen; Miller, Jon M.; Wijnands, Rudy

2016-12-01

Crustal cooling of accretion-heated neutron stars provides insight into the stellar interior of neutron stars. The neutron star X-ray transient, KS 1731-260, was in outburst for 12.5 years before returning to quiescence in 2001. We have monitored the cooling of this source since then through Chandra and XMM-Newton observations. Here we present a 150 ks Chandra observation of KS 1731-260 taken in 2015 August, about 14.5 years into quiescence and 6 years after the previous observation. We find that the neutron star surface temperature is consistent with the previous observation, suggesting that crustal cooling has likely stopped and the crust has reached thermal equilibrium with the core. Using a theoretical crust thermal evolution code, we fit the observed cooling curves and constrain the core temperature (T c = 9.35 ± 0.25 × 107 K), composition (Q {}{imp}={4.4}-0.5+2.2), and level of extra shallow heating required (Q sh = 1.36 ± 0.18 MeV/nucleon). We find that the presence of a low thermal conductivity layer, as expected from nuclear pasta, is not required to fit the cooling curve well, but cannot be excluded either.

The thermal state of KS 1731-260 after 14.5 years in quiescence

NASA Astrophysics Data System (ADS)

Merritt, R.; Cackett, E.; Brown, E.; Page, D.; Cumming, A.; Degenaar, N.; Deibel, A.; Homan, J.; Miller, J.; Wijnands, R.

2017-10-01

Crustal cooling of accretion-heated neutron stars provides insight into the stellar interior of neutron stars. The neutron star X-ray transient, KS 1731-260, was in outburst for 12.5 years before returning to quiescence in 2001. We have monitored the cooling of this source since then through Chandra and XMM-Newton observations. Here, we present a 150 ks Chandra observation of KS 1731-260 taken in August 2015, about 14.5 years into quiescence, and 6 years after the previous observation. We find that the neutron star surface temperature is consistent with the previous observation, suggesting that crustal cooling has likely stopped and the crust has reached thermal equilibrium with the core. Using a theoretical crust thermal evolution code, we fit the observed cooling curves and constrain the core temperature (T_c = 9.35±0.25×10^7 K), composition (Q_{imp} = 4.4^{+2.2}_{-0.5}) and level of extra shallow heating required (Q_{sh} = 1.36±0.18 MeV/nucleon). We find that the presence of a low thermal conductivity layer, as expected from nuclear pasta, is not required to fit the cooling curve well, but cannot be excluded either.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Melting and vibrational properties of planetary materials under deep Earth conditions

NASA Astrophysics Data System (ADS)

Jackson, Jennifer

2013-06-01

The large chemical, density, and dynamical contrasts associated with the juxtaposition of a liquid iron-dominant alloy and silicates at Earth's core-mantle boundary (CMB) are associated with a rich range of complex seismological features. For example, seismic heterogeneity at this boundary includes small patches of anomalously low sound velocities, called ultralow-velocity zones. Their small size (5 to 40 km thick) and depth (about 2800 km) present unique challenges for seismic characterization and geochemical interpretation. In this contribution, we will present recent nuclear resonant inelastic x-ray scattering measurements on iron-bearing silicates, oxides, and metals, and their application towards our understanding of Earth's interior. Specifically, we will present measurements on silicates and oxide minerals that are important in Earth's upper and lower mantles, as well as iron to over 1 megabar in pressure. The nuclear resonant inelastic x-ray scattering method provides specific vibrational information, e.g., the phonon density of states, and in combination with compression data permits the determination of sound velocities and other vibrational information under high pressure and high temperature. For example, accurate determination of the sound velocities and density of chemically complex Earth materials is essential for understanding the distribution and behavior of minerals and iron-alloys with depth. The high statistical quality of the data in combination with high energy resolution and a small x-ray focus size permit accurate evaluation of the vibrational-related quantities of iron-bearing Earth materials as a function of pressure, such as the Grüneisen parameter, thermal pressure, sound velocities, and iron isotope fractionation quantities. Finally, we will present a novel method detecting the solid-liquid phase boundary of compressed iron at high temperatures using synchrotron Mössbauer spectroscopy. Our approach is unique because the dynamics of the iron atoms are monitored. This process is described by the Lamb-Mössbauer factor, which is related to the mean-square displacement of the iron atoms. We will discuss the implications of our results as they relate to Earth's core and core-mantle boundary regions.

The host galaxy/AGN connection in nearby early-type galaxies. Is there a miniature radio-galaxy in every "core" galaxy?

NASA Astrophysics Data System (ADS)

Balmaverde, B.; Capetti, A.

2006-02-01

This is the second of a series of three papers exploring the connection between the multiwavelength properties of AGN in nearby early-type galaxies and the characteristics of their hosts. We selected two samples with 5 GHz VLA radio flux measurements down to 1 mJy, reaching levels of radio luminosity as low as 1036 erg s-1. In Paper I we presented a study of the surface brightness profiles for the 65 objects with available archival HST images out of the 116 radio-detected galaxies. We classified early-type galaxies into "core" and "power-law" galaxies, discriminating on the basis of the slope of their nuclear brightness profiles, following the Nukers scheme. Here we focus on the 29 core galaxies (hereafter CoreG). We used HST and Chandra data to isolate their optical and X-ray nuclear emission. The CoreG invariably host radio-loud nuclei, with an average radio-loudness parameter of Log R = L5 {GHz} / LB ˜ 3.6. The optical and X-ray nuclear luminosities correlate with the radio-core power, smoothly extending the analogous correlations already found for low luminosity radio-galaxies (LLRG) toward even lower power, by a factor of ˜ 1000, covering a combined range of 6 orders of magnitude. This supports the interpretation of a common non-thermal origin of the nuclear emission also for CoreG. The luminosities of the nuclear sources, most likely dominated by jet emission, set firm upper limits, as low as L/L_Edd ˜ 10-9 in both the optical and X-ray band, on any emission from the accretion process. The similarity of CoreG and LLRG when considering the distributions host galaxies luminosities and black hole masses, as well as of the surface brightness profiles, indicates that they are drawn from the same population of early-type galaxies. LLRG represent only the tip of the iceberg associated with (relatively) high activity levels, with CoreG forming the bulk of the population. We do not find any relationship between radio-power and black hole mass. A minimum black hole mass of M_BH = 108 M⊙ is apparently associated with the radio-loud nuclei in both CoreG and LLRG, but this effect must be tested on a sample of less luminous galaxies, likely to host smaller black holes. In the unifying model for BL Lacs and radio-galaxies, CoreG likely represent the counterparts of the large population of low luminosity BL Lac now emerging from the surveys at low radio flux limits. This suggests the presence of relativistic jets also in these quasi-quiescent early-type "core" galaxies.

A Novel In-Beam Delayed Neutron Counting Technique for Characterization of Special Nuclear Materials

NASA Astrophysics Data System (ADS)

Bentoumi, G.; Rogge, R. B.; Andrews, M. T.; Corcoran, E. C.; Dimayuga, I.; Kelly, D. G.; Li, L.; Sur, B.

2016-12-01

A delayed neutron counting (DNC) system, where the sample to be analyzed remains stationary in a thermal neutron beam outside of the reactor, has been developed at the National Research Universal (NRU) reactor of the Canadian Nuclear Laboratories (CNL) at Chalk River. The new in-beam DNC is a novel approach for non-destructive characterization of special nuclear materials (SNM) that could enable identification and quantification of fissile isotopes within a large and shielded sample. Despite the orders of magnitude reduction in neutron flux, the in-beam DNC method can be as informative as the conventional in-core DNC for most cases while offering practical advantages and mitigated risk when dealing with large radioactive samples of unknown origin. This paper addresses (1) the qualification of in-beam DNC using a monochromatic thermal neutron beam in conjunction with a proven counting apparatus designed originally for in-core DNC, and (2) application of in-beam DNC to an examination of large sealed capsules containing unknown radioactive materials. Initial results showed that the in-beam DNC setup permits non-destructive analysis of bulky and gamma shielded samples. The method does not lend itself to trace analysis, and at best could only reveal the presence of a few milligrams of 235U via the assay of in-beam DNC total counts. Through analysis of DNC count rates, the technique could be used in combination with other neutron or gamma techniques to quantify isotopes present within samples.

Transport Properties of Earth's Core

NASA Astrophysics Data System (ADS)

Cohen, R. E.; Zhang, P.; Xu, J.

2016-12-01

One of the most important parameters governing the original heat that drives all processes in the Earth is the thermal conductivity of Earth's core. Heat is transferred through the core by convection and conduction, and the convective component provides energy to drive the geodynamo. Sha and Cohen (2011) found that the electrical conductivity of solid hcp-iron was much higher than had been assumed by geophysicists, based on electronic structure computations for electron-phonon scattering (e-p) within density functional theory [1]. Thermal conductivity is related to electrical conductivity through the empirical Wiedmann-Franz law of 1853 [2]. Pozzo et al. [3] found that the high electrical conductivity of liquid iron alloys was too high for conventional dynamo models to work—there simply is not enough energy, so O'Rourke and Stevenson proposed a model driven by participation of Mg from the core [4], supported by recent experients [5]. Recent measurements by Ohta et al. show even lower resistivities than predicted by DFT e-p, and invoked a saturation model to account for this, [6] whereas, Konopkova et al. found thermal conductivities consistent with earlier geophysical estimates. [7] We are using first-principles methods, including dynamical mean field theory for electron-electron scattering, and highly converged e-p computations, and find evidence for strong anisotropy in solid hcp-Fe that may help explain some experimental results. The current status of the field will be discussed along with our recent results. This work is supported by the ERC Advanced grant ToMCaT, the NSF, and the Carnegie Institution for Science.[1] X. Sha and R. E. Cohen, J.Phys.: Condens.Matter 23, 075401 (2011).[2] R. Franz and G. Wiedemann, Annalen Physik 165, 497 (1853).[3] M. Pozzo, C. Davies, D. Gubbins, and D. Alfe, Nature 485, 355 (2012).[4] J. G. O'Rourke and D. J. Stevenson, Nature 529, 387 (2016).[5] J. Badro, J. Siebert, and F. Nimmo, Nature (2016).[6] K. Ohta, Y. Kuwayama, K. Hirose, K. Shimizu, and Y. Ohishi, Nature 534, 95 (2016).[7] Z. Konopkova, R. S. McWilliams, N. Gomez-Perez, and A. F. Goncharov, Nature 534, 99 (2016).

Absolute efficiency calibration of 6LiF-based solid state thermal neutron detectors

NASA Astrophysics Data System (ADS)

Finocchiaro, Paolo; Cosentino, Luigi; Lo Meo, Sergio; Nolte, Ralf; Radeck, Desiree

2018-03-01

The demand for new thermal neutron detectors as an alternative to 3He tubes in research, industrial, safety and homeland security applications, is growing. These needs have triggered research and development activities about new generations of thermal neutron detectors, characterized by reasonable efficiency and gamma rejection comparable to 3He tubes. In this paper we show the state of the art of a promising low-cost technique, based on commercial solid state silicon detectors coupled with thin neutron converter layers of 6LiF deposited onto carbon fiber substrates. A few configurations were studied with the GEANT4 simulation code, and the intrinsic efficiency of the corresponding detectors was calibrated at the PTB Thermal Neutron Calibration Facility. The results show that the measured intrinsic detection efficiency is well reproduced by the simulations, therefore validating the simulation tool in view of new designs. These neutron detectors have also been tested at neutron beam facilities like ISIS (Rutherford Appleton Laboratory, UK) and n_TOF (CERN) where a few samples are already in operation for beam flux and 2D profile measurements. Forthcoming applications are foreseen for the online monitoring of spent nuclear fuel casks in interim storage sites.

Addition polymers from 1,4,5,8-tetrahydro-1,4;5,8-diepoxyanthracene and Bis-dienes. 2: Evidence for thermal dehydration occurring in the cure process

NASA Technical Reports Server (NTRS)

Meador, Mary Ann B.; Olshavsky, Michael A.; Meador, Michael A.; Ahn, Myong-Ku

1988-01-01

Diels-Alder cycloaddition copolymers from 1,4,5,8-tetrahydro-1,4;5,8-diepoxyanthracene and anthracene end-capped polyimide oligomers appear, by thermogravimetric analysis (TGA), to undergo dehydration at elevated temperatures. This would produce thermally stable pentiptycene units along the polymer backbone, and render the polymers incapable of unzipping through a retro-Diels-Alder pathway. High resolution solid 13C nuclear magnetic resonance (NMR) of one formulation of the polymer system before and after heating at elevated temperatures, shows this to indeed be the case. NMR spectra of solid samples of the polymer before and after heating correlated well with those of the parent pentiptycene model compound before and after acid-catalyzed dehydration. Isothermal gravimetric analyses and viscosities of the polymer before and after heat treatment support dehydration as a mechanism for the cure reaction.

A Nuclear Reactor and Chemical Processing Design for Production of Molybdenum-99 with Crystalline Uranyl Nitrate Hexahydrate Fuel

NASA Astrophysics Data System (ADS)

Stange, Gary Michael

Medical radioisotopes are used in tens of millions of procedures every year to detect and image a wide variety of maladies and conditions in the human body. The most widely-used diagnostic radioisotope is technetium-99m, a metastable isomer of technetium-99 that is generated by the radioactive decay of molybdenum-99. For a number of reasons, the supply of molybdenum-99 has become unreliable and the techniques used to produce it have become unattractive. This has spurred the investigation of new technologies that avoid the use of highly enriched uranium to produce molybdenum-99 in the United States, where approximately half of the demand originates. The first goal of this research is to develop a critical nuclear reactor design powered by solid, discrete pins of low enriched uranium. Analyses of single-pin heat transfer and whole-core neutronics are performed to determine the required specifications. Molybdenum-99 is produced directly in the fuel of this reactor and then extracted through a series of chemical processing steps. After this extraction, the fuel is left in an aqueous state. The second goal of this research is to describe a process by which the uranium may be recovered from this spent fuel solution and reconstituted into the original fuel form. Fuel recovery is achieved through a crystallization step that generates solid uranyl nitrate hexahydrate while leaving the majority of fission products and transuranic isotopes in solution. This report provides background information on molybdenum-99 production and crystallization chemistry. The previously unknown thermal conductivity of the fuel material is measured. Following this is a description of the modeling and calculations used to develop a reactor concept. The operational characteristics of the reactor core model are analyzed and reported. Uranyl nitrate crystallization experiments have also been conducted, and the results of this work are presented here. Finally, a process flow scheme for uranium recovery is examined, in part qualitatively and in part quantitatively, based upon the preceding data garnered through literature review, modeling, and experimentation. The sum of this research is meant to allow for a complete understanding of the process flow, from the beginning of one production cycle to the beginning of another.

Synthesis of Multicolor Core/Shell NaLuF4:Yb3+/Ln3+@CaF2 Upconversion Nanocrystals

PubMed Central

Li, Hui; Hao, Shuwei; Yang, Chunhui; Chen, Guanying

2017-01-01

The ability to synthesize high-quality hierarchical core/shell nanocrystals from an efficient host lattice is important to realize efficacious photon upconversion for applications ranging from bioimaging to solar cells. Here, we describe a strategy to fabricate multicolor core @ shell α-NaLuF4:Yb3+/Ln3+@CaF2 (Ln = Er, Ho, Tm) upconversion nanocrystals (UCNCs) based on the newly established host lattice of sodium lutetium fluoride (NaLuF4). We exploited the liquid-solid-solution method to synthesize the NaLuF4 core of pure cubic phase and the thermal decomposition approach to expitaxially grow the calcium fluoride (CaF2) shell onto the core UCNCs, yielding cubic core/shell nanocrystals with a size of 15.6 ± 1.2 nm (the core ~9 ± 0.9 nm, the shell ~3.3 ± 0.3 nm). We showed that those core/shell UCNCs could emit activator-defined multicolor emissions up to about 772 times more efficient than the core nanocrystals due to effective suppression of surface-related quenching effects. Our results provide a new paradigm on heterogeneous core/shell structure for enhanced multicolor upconversion photoluminescence from colloidal nanocrystals. PMID:28336867

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Encapsulated nano-heat-sinks for thermal management of heterogeneous chemical reactions.

PubMed

Zhang, Minghui; Hong, Yan; Ding, Shujiang; Hu, Jianjun; Fan, Yunxiao; Voevodin, Andrey A; Su, Ming

2010-12-01

This paper describes a new way to control temperatures of heterogeneous exothermic reactions such as heterogeneous catalytic reaction and polymerization by using encapsulated nanoparticles of phase change materials as thermally functional additives. Silica-encapsulated indium nanoparticles and silica encapsulated paraffin nanoparticles are used to absorb heat released in catalytic reaction and to mitigate gel effect of polymerization, respectively. The local hot spots that are induced by non-homogenous catalyst packing, reactant concentration fluctuation, and abrupt change of polymerization rate lead to solid to liquid phase change of nanoparticle cores so as to avoid thermal runaway by converting energies from exothermic reactions to latent heat of fusion. By quenching local hot spots at initial stage, reaction rates do not rise significantly because the thermal energy produced in reaction is isothermally removed. Nanoparticles of phase change materials will open a new dimension for thermal management of exothermic reactions to quench local hot spots, prevent thermal runaway of reaction, and change product distribution.

The ab initio simulation of the Earth's core.

PubMed

Alfè, D; Gillan, M J; Vocadlo, L; Brodholt, J; Price, G D

2002-06-15

The Earth has a liquid outer and solid inner core. It is predominantly composed of Fe, alloyed with small amounts of light elements, such as S, O and Si. The detailed chemical and thermal structure of the core is poorly constrained, and it is difficult to perform experiments to establish the properties of core-forming phases at the pressures (ca. 300 GPa) and temperatures (ca. 5000-6000 K) to be found in the core. Here we present some major advances that have been made in using quantum mechanical methods to simulate the high-P/T properties of Fe alloys, which have been made possible by recent developments in high-performance computing. Specifically, we outline how we have calculated the Gibbs free energies of the crystalline and liquid forms of Fe alloys, and so conclude that the inner core of the Earth is composed of hexagonal close packed Fe containing ca. 8.5% S (or Si) and 0.2% O in equilibrium at 5600 K at the boundary between the inner and outer cores with a liquid Fe containing ca. 10% S (or Si) and 8% O.

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

Analysis of failed nuclear plant components

NASA Astrophysics Data System (ADS)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

Impact of intermediate and high energy nuclear data on the neutronic safety parameters of MYRRHA accelerator driven system

NASA Astrophysics Data System (ADS)

Stankovskiy, Alexey; Çelik, Yurdunaz; Eynde, Gert Van den

2017-09-01

Perturbation of external neutron source can cause significant local power changes transformed into undesired safety-related events in an accelerator driven system. Therefore for the accurate design of MYRRHA sub-critical core it is important to evaluate the uncertainty of power responses caused by the uncertainties in nuclear reaction models describing the particle transport from primary proton energy down to the evaluated nuclear data table range. The calculations with a set of models resulted in quite low uncertainty on the local power caused by significant perturbation of primary neutron yield from proton interactions with lead and bismuth isotopes. The considered accidental event of prescribed proton beam shape loss causes drastic increase in local power but does not practically change the total core thermal power making this effect difficult to detect. In the same time the results demonstrate a correlation between perturbed local power responses in normal operation and misaligned beam conditions indicating that generation of covariance data for proton and neutron induced neutron multiplicities for lead and bismuth isotopes is needed to obtain reliable uncertainties for local power responses.

Molecular dynamics for dense matter

NASA Astrophysics Data System (ADS)

Maruyama, Toshiki; Watanabe, Gentaro; Chiba, Satoshi

2012-08-01

We review a molecular dynamics method for nucleon many-body systems called quantum molecular dynamics (QMD), and our studies using this method. These studies address the structure and the dynamics of nuclear matter relevant to neutron star crusts, supernova cores, and heavy-ion collisions. A key advantage of QMD is that we can study dynamical processes of nucleon many-body systems without any assumptions about the nuclear structure. First, we focus on the inhomogeneous structures of low-density nuclear matter consisting not only of spherical nuclei but also of nuclear "pasta", i.e., rod-like and slab-like nuclei. We show that pasta phases can appear in the ground and equilibrium states of nuclear matter without assuming nuclear shape. Next, we show our simulation of compression of nuclear matter which corresponds to the collapsing stage of supernovae. With the increase in density, a crystalline solid of spherical nuclei changes to a triangular lattice of rods by connecting neighboring nuclei. Finally, we discuss fragment formation in expanding nuclear matter. Our results suggest that a generally accepted scenario based on the liquid-gas phase transition is not plausible at lower temperatures.

Physical particularities of nuclear reactors using heavy moderators of neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N.

2016-12-15

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program packagemore » for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.« less

Dual-mode, high energy utilization system concept for mars missions

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2000-01-01

This paper describes a dual-mode, high energy utilization system concept based on the Pellet Bed Reactor (PeBR) to support future manned missions to Mars. The system uses proven Closed Brayton Cycle (CBC) engines to partially convert the reactor thermal power to electricity. The electric power generated is kept the same during the propulsion and the power modes, but the reactor thermal power in the former could be several times higher, while maintaining the reactor temperatures almost constant. During the propulsion mode, the electric power of the system, minus ~1-5 kWe for house keeping, is used to operate a Variable Specific Impulse Magnetoplasma Rocket (VASIMR). In addition, the reactor thermal power, plus more than 85% of the head load of the CBC engine radiators, are used to heat hydrogen. The hot hydrogen is mixed with the high temperature plasma in a VASIMR to provide both high thrust and Isp>35,000 N.s/kg, reducing the travel time to Mars to about 3 months. The electric power also supports surface exploration of Mars. The fuel temperature and the inlet temperatures of the He-Xe working fluid to the nuclear reactor core and the CBC turbine are maintained almost constant during both the propulsion and power modes to minimize thermal stresses. Also, the exit temperature of the He-Xe from the reactor core is kept at least 200 K below the maximum fuel design temperature. The present system has no single point failure and could be tested fully assembled in a ground facility using electric heaters in place of the nuclear reactor. Operation and design parameters of a 40-kWe prototype are presented and discussed to illustrate the operation and design principles of the proposed system. .

The physics of solid-state neutron detector materials and geometries.

PubMed

Caruso, A N

2010-11-10

Detection of neutrons, at high total efficiency, with greater resolution in kinetic energy, time and/or real-space position, is fundamental to the advance of subfields within nuclear medicine, high-energy physics, non-proliferation of special nuclear materials, astrophysics, structural biology and chemistry, magnetism and nuclear energy. Clever indirect-conversion geometries, interaction/transport calculations and modern processing methods for silicon and gallium arsenide allow for the realization of moderate- to high-efficiency neutron detectors as a result of low defect concentrations, tuned reaction product ranges, enhanced effective omnidirectional cross sections and reduced electron-hole pair recombination from more physically abrupt and electronically engineered interfaces. Conversely, semiconductors with high neutron cross sections and unique transduction mechanisms capable of achieving very high total efficiency are gaining greater recognition despite the relative immaturity of their growth, lithographic processing and electronic structure understanding. This review focuses on advances and challenges in charged-particle-based device geometries, materials and associated mechanisms for direct and indirect transduction of thermal to fast neutrons within the context of application. Calorimetry- and radioluminescence-based intermediate processes in the solid state are not included.

MgCoAl and NiCoAl LDHs synthesized by the hydrothermal urea hydrolysis method: Structural characterization and thermal decomposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chagas, L.H., E-mail: lhchagas-prometro@inmetro.gov.br; Instituto Nacional de Metrologia Qualidade e Tecnologia, Divisão de Metrologia de Materiais, 25250-020 Duque de Caxias, RJ; De Carvalho, G.S.G.

Highlights: • We synthesized MgCoAl and NiCoAl LDHs by the urea hydrolysis method. • Aluminum rich and crystalline materials have been formed. • The calcination of the LDHs generated mixed oxides with high surface areas. - Abstract: Layered double hydroxides (LDHs) with Mg/Co/Al and Ni/Co/Al were synthesized for the first time by the urea hydrolysis method. The experimental conditions promoted aluminum rich and crystalline materials. The formation of LDHs was investigated by powder X-ray diffraction (XRD), chemical analysis, solid state nuclear magnetic resonance with magic angle spinning ({sup 27}Al-MAS-NMR), simultaneous thermogravimetric/differential thermal analysis (TGA/DTA), FTIR spectroscopy, scanning electron microscopy (SEM),more » and N{sub 2} adsorption–desorption experiments. A single phase corresponding to LDH could be obtained in all the investigated compositions. Thermal calcination of these LDHs at 500 °C resulted in the formation of solid solutions in which Al{sup 3+} was dissolved. All the calcined materials have rock-salt like structures and high surface areas.« less

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

DOE Office of Scientific and Technical Information (OSTI.GOV)

farahani, A.A.; Corradini, M.L.

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for themore » foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.« less

The thermal evolution and dynamo generation of Mercury with an Fe-Si core

NASA Astrophysics Data System (ADS)

Knibbe, Jurrien

2017-04-01

The present day partially liquid (as opposed to fully solidified) Fe-rich core of Mercury is traditionally explained by assuming a substantial amount of S to be present in the core (e.g. Grott et al., 2011), because S lowers the core's melting temperature. However, this assumption has problematic implications: Mercury's large Fe-rich core and measured low FeO surface content are indicative of an oxygen poor bulk composition, which is consistent with the volatile-poor material that is expected to have condensed from the solar nebula close to the Sun. In contrast, S is a moderately volatile element. Combined with the high S content of Mercury's crust and (likely) mantle, as indicated by the measured high S/Si surface fraction, the resulting high planetary S abundance is difficult to reconcile with a volatile poor origin of the planet. Additionally, the observed low magnetic field strength is most easily explained if compositional buoyancy fluxes are absent [Manglik et al., 2010], yet such fluxes are produced upon solidifying a pure Fe inner core from Fe-S liquid. Alternatively, both Mercury's high S/Si and Mg/Si surface ratios (Nittler et al., 2011) may indicate that a siderophile fractionation of Si and lithophile fractionation of S took place during Mercury's core-mantle differentiation. This fractionation behaviour of these elements is supported by metal/silicate partitioning experiments that have been performed at the low oxygen conditions inferred for Mercury [e.g. Chabot et al., 2014]. Mercury's bulk composition, in terms of S/Si and Fe/Si ratios, would also approach that of meteorites that are considered as potential building blocks of the planet if the core is Si-rich and S-poor. Here we simulate the thermal evolution of Mercury with an Fe-Si core. Results show that an Fe-Si core can remain largely molten until present, without the need for S. An Fe-Si core also has interesting implications for Mercury's core-convection regime and magnetic field generation. The non-preferential Si fractionation between solid and liquid metal does not produce a compositional gradient, such that compositional buoyancy fluxes are negligible. Additionally, thermally driven core convection is more efficient as a result of a high latent heat release upon solidifying Si-rich metal. Implications of this scenario for Mercury's magnetic field strength and geometry need to be further examined.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

Thermophysical properties of U 3 Si 2 to 1773K

DOE PAGES

White, Joshua Taylor; Nelson, Andrew Thomas; Dunwoody, John Tyler; ...

2015-05-08

Use of U 3Si 2 in nuclear reactors requires accurate thermophysical property data to capture heat transfer within the core. Compilation of the limited previous research efforts focused on the most critical property, thermal conductivity, reveals extensive disagreement. Assessment of this data is challenged by the fact that the critical structural and chemical details of the material used to provide historic data is either absent or confirms the presence of significant impurity phases. This study was initiated to fabricate high purity U 3Si 2 to quantify the coefficient of thermal expansion, heat capacity, thermal diffusivity, and thermal conductivity from roommore » temperature to 1773 K. Here, the datasets provided in this manuscript will facilitate more detailed fuel performance modeling to assess both current and proposed reactor designs that incorporate U 3Si 2.« less

NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

DOEpatents

Mills, F.T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

Neutronic Reactor Design to Reduce Neutron Loss

DOEpatents

Miles, F. T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

Characterization of exposures to nanoscale particles and fibers during solid core drilling of hybrid carbon nanotube advanced composites.

PubMed

Bello, Dhimiter; Wardle, Brian L; Zhang, Jie; Yamamoto, Namiko; Santeufemio, Christopher; Hallock, Marilyn; Virji, M Abbas

2010-01-01

This work investigated exposures to nanoparticles and nanofibers during solid core drilling of two types of advanced carbon nanotube (CNT)-hybrid composites: (1) reinforced plastic hybrid laminates (alumina fibers and CNT); and (2) graphite-epoxy composites (carbon fibers and CNT). Multiple real-time instruments were used to characterize the size distribution (5.6 nm to 20 microm), number and mass concentration, particle-bound polyaromatic hydrocarbons (b-PAHs), and surface area of airborne particles at the source and breathing zone. Time-integrated samples included grids for electron microscopy characterization of particle morphology and size resolved (2 nm to 20 microm) samples for the quantification of metals. Several new important findings herein include generation of airborne clusters of CNTs not seen during saw-cutting of similar composites, fewer nanofibers and respirable fibers released, similarly high exposures to nanoparticles with less dependence on the composite thickness, and ultrafine (< 5 nm) aerosol originating from thermal degradation of the composite material.

Phase-coherent engineering of electronic heat currents with a Josephson modulator

NASA Astrophysics Data System (ADS)

Fornieri, Antonio; Blanc, Christophe; Bosisio, Riccardo; D'Ambrosio, Sophie; Giazotto, Francesco

In this contribution we report the realization of the first balanced Josephson heat modulator designed to offer full control at the nanoscale over the phase-coherent component of electronic thermal currents. The ability to master the amount of heat transferred through two tunnel-coupled superconductors by tuning their phase difference is the core of coherent caloritronics, and is expected to be a key tool in a number of nanoscience fields, including solid state cooling, thermal isolation, radiation detection, quantum information and thermal logic. Our device provides magnetic-flux-dependent temperature modulations up to 40 mK in amplitude with a maximum of the flux-to-temperature transfer coefficient reaching 200 mK per flux quantum at a bath temperature of 25 mK. Foremost, it demonstrates the exact correspondence in the phase-engineering of charge and heat currents, breaking ground for advanced caloritronic nanodevices such as thermal splitters, heat pumps and time-dependent electronic engines.

Application of gaseous core reactors for transmutation of nuclear waste

NASA Technical Reports Server (NTRS)

Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

1976-01-01

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

A Thermally Re-mendable Cross-Linked Polymeric Material

NASA Astrophysics Data System (ADS)

Chen, Xiangxu; Dam, Matheus A.; Ono, Kanji; Mal, Ajit; Shen, Hongbin; Nutt, Steven R.; Sheran, Kevin; Wudl, Fred

2002-03-01

We have developed a transparent organic polymeric material that can repeatedly mend or ``re-mend'' itself under mild conditions. The material is a tough solid at room temperature and below with mechanical properties equaling those of commercial epoxy resins. At temperatures above 120°C, approximately 30% (as determined by solid-state nuclear magnetic resonance spectroscopy) of ``intermonomer'' linkages disconnect but then reconnect upon cooling, This process is fully reversible and can be used to restore a fractured part of the polymer multiple times, and it does not require additional ingredients such as a catalyst, additional monomer, or special surface treatment of the fractured interface.

Low-Impact Space Weather Sensors and the U.S. National Security Spacecraft

DTIC Science & Technology

2016-09-01

aircraft. Even the CIA’s supersonic and stealthy A-12 Oxcart and the Air Force’s SR-71 were vulnerable by the time they became operational. The...thrusters that expel the energy providing thrust. Multiple sources of energy can be used for propellant, including solid and liquid fueled thrusters...dynamic space weather. At its core, this great ball of gas produces significant energy through nuclear fusion that converts hydrogen to helium, the two

Development of a New 47-Group Library for the CASL Neutronics Simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Williams, Mark L; Wiarda, Dorothea

The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACTmore » library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.« less

AGC-2 Irradiation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohrbaugh, David Thomas; Windes, William; Swank, W. David

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a completemore » properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphites is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.« less

Fast round-trip Mars trajectories

NASA Technical Reports Server (NTRS)

Wilson, Sam

1990-01-01

This paper is concerned with the effect of limiting the overall duration or else the one-way flight time of a round trip to Mars, as reflected in the sum of impulsive velocity increments required of the spacecraft propulsion system. Ignition-to-burnout mass ratios for a hypothetical single stage spacecraft, obtained from the rocket equation by combining these delta-V sums with appropriate values of specific impulse, are used to evaluate the relative effectiveness of four high-thrust propulsion alternatives. If the flight crew goes to the surface of Mars and stays there for the duration of their stopover, it is much cheaper (in terms of delta-V) to minimize their zero-g exposure by limiting the interplanetary transit time of a conjunction-class mission (round trip time = 800-1000 days, Mars stopover = 450-700 days) than to impose the same limit on an opposition-class mission (round trip time less than 600 days, stopover = 40 days). Using solid-core nuclear thermal propulsion to fly a conjunction-class mission, for a moderate mass penalty the interplanetary transit time (each way) probably could be limited to something in the range of 4 to 6 months, depending on the launch year.

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

Indirect measurement of the solid/liquid interface using the minimization technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, H.; Chun, M.

1985-11-01

The phenomenon of solidification of a flowing fluid in a vertical tube is closely related to the relocation dynamics of molten nuclear fuels in hypothetical core-disruptive accidents of a liquid-metal fast breeder reactor. The knowledge of the transient shape and the position of the liquid/solid interface is of practical importance in analysis of phase change processes. Sparrow and Broadbent directly measured the solid liquid interface via experiments, whereas Viskanta observed the solid/liquid interface motion via a photographic method. In this paper, a new method to predict the transient position of the solid/liquid interface is developed. This method is based onmore » the minimization technique. To use this method one needs the temperature of the wall on which the phase change is to take place. The new technique is useful, in particular, for the case of inward solidification of a flowing fluid in a tube where direct measurement of the solid/liquid interface is not possible, whereas the tube wall temperature measurement is relatively easy.« less

Characterization of the Kinetics of NF3-Fluorination of NpO2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Casella, Andrew M.; Scheele, Randall D.; McNamara, Bruce K.

2015-12-23

The exploitation of selected actinide and fission product fluoride volatilities has long been considered as a potentially attractive compact method for recycling used nuclear fuels to avoid generating the large volumes of radioactive waste arising from aqueous reprocessing [1-7]. The most developed process uses the aggressive and hazardous fluorinating agents hydrogen fluoride (HF) and/or molecular fluorine (F2) at high temperatures to volatilize the greatest fraction of the used nuclear fuel into a single gas stream. The volatilized fluorides are subsequently separated using a series of fractionation and condensation columns to recover the valuable fuel constituents and fission products. In pursuitmore » of a safer and less complicated approach, we investigated an alternative fluoride volatility-based process using the less hazardous fluorinating agent nitrogen trifluoride (NF3) and leveraging its less aggressive nature to selectively evolve fission product and actinide fluorides from the solid phase based on their reaction temperatures into a single recycle stream [8-15]. In this approach, successive isothermal treatments using NF3 will first evolve the more thermally susceptible used nuclear fuel constituents leaving the other constituents in the residual solids until subsequent isothermal temperature treatments cause these others to volatilize. During investigation of this process, individual neat used fuel components were treated with isothermal NF3 in an attempt to characterize the kinetics of each fluorination reaction to provide input into the design of a new volatile fluoride separations approach. In these directed investigations, complex behavior was observed between NF3 and certain solid reactants such as the actinide oxides of uranium, plutonium, and neptunium. Given the similar thermal reaction susceptibilities of neptunium oxide (NpO2) and uranium dioxide (UO2) and the importance of Np and U, we initially focused our efforts on determining the reaction kinetic parameters for NpO2. Characterizing the NF3 fluorination of NpO2 using established models for gas-solid reactions [16] proved unsuccessful so we developed a series of successive fundamental reaction mechanisms to characterize the observed successive fluorination reactions leading to production of the volatile neptunium hexafluoride (NpF6).« less

Nuclear and Particle Physics Simulations: The Consortium of Upper-Level Physics Software

NASA Astrophysics Data System (ADS)

Bigelow, Roberta; Moloney, Michael J.; Philpott, John; Rothberg, Joseph

1995-06-01

The Consortium for Upper Level Physics Software (CUPS) has developed a comprehensive series of Nine Book/Software packages that Wiley will publish in FY `95 and `96. CUPS is an international group of 27 physicists, all with extensive backgrounds in the research, teaching, and development of instructional software. The project is being supported by the National Science Foundation (PHY-9014548), and it has received other support from the IBM Corp., Apple Computer Corp., and George Mason University. The Simulations being developed are: Astrophysics, Classical Mechanics, Electricity & Magnetism, Modern Physics, Nuclear and Particle Physics, Quantum Mechanics, Solid State, Thermal and Statistical, and Wave and Optics.

The investigation of the effect of thermal treatment on bentonites from Turkey with Fourier transform infrared and solid state nuclear magnetic resonance spectroscopic methods.

PubMed

Erdoğan Alver, Burcu; Alver, Ozgür

2012-08-01

There is a great deal of interest in the building industry in burned clays for production of building materials. Therefore, the effect of heat treatment on natural bentonite from Turkey was investigated by Fourier transform infrared (FT-IR) between the region of 4000-400cm(-1) and (29)Si, (27)Al magic angle spinning nuclear magnetic resonance (MAS NMR) measurement techniques at various temperatures between 200 and 700°C for 2h. The structural changes were also investigated upon heat treatment. Copyright © 2012 Elsevier B.V. All rights reserved.

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

Technology Solutions Case Study: Durable Interior Foundation Insulation Retrofits for Cold Climates, Cloquet, Minnesota

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Thermal and moisture problems in existing basements create a unique challenge as the exterior face of the wall is not easily or inexpensively accessible. This approach by the NorthernSTAR Building America Partnership team addresses thermal and moisture management from the interior face of the wall without disturbing the exterior soil and landscaping. It is effective at reducing energy loss through the wall principally during the heating season. The team conducted experiments at the Cloquet Residential Research Facility to test the heat and moisture performance of four hollow masonry block wall systems and two rim-joist systems. These systems were retrofitted withmore » interior insulation in compliance with the 2012 IECC. The research showed for the first time that, for masonry block walls in a cold climate, a solid bond beam or equivalent provides adequate resistance to moisture transport from a hollow core to the rim-joist cavity. Thus, a solid top course is a minimum requirement for an interior retrofit insulation system.« less

Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

2017-01-15

This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches andmore » time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.« less

Experiments on Lunar Core Composition: Phase Equilibrium Analysis of A Multi-Element (Fe-Ni-S-C) System

NASA Technical Reports Server (NTRS)

Go, B. M.; Righter, K.; Danielson, L.; Pando, K.

2015-01-01

Previous geochemical and geophysical experiments have proposed the presence of a small, metallic lunar core, but its composition is still being investigated. Knowledge of core composition can have a significant effect on understanding the thermal history of the Moon, the conditions surrounding the liquid-solid or liquid-liquid field, and siderophile element partitioning between mantle and core. However, experiments on complex bulk core compositions are very limited. One limitation comes from numerous studies that have only considered two or three element systems such as Fe-S or Fe-C, which do not supply a comprehensive understanding for complex systems such as Fe-Ni-S-Si-C. Recent geophysical data suggests the presence of up to 6% lighter elements. Reassessments of Apollo seismological analyses and samples have also shown the need to acquire more data for a broader range of pressures, temperatures, and compositions. This study considers a complex multi-element system (Fe-Ni-S-C) for a relevant pressure and temperature range to the Moon's core conditions.

Magneto thermal conductivity of superconducting Nb with intermediate level of impurity

DOE Office of Scientific and Technical Information (OSTI.GOV)

L.S. Sharath Chandra, M.K. Chattopadhyay, S.B. Roy, V.C. Sahni, G.R. Myneni

2012-03-01

Niobium materials with intermediate purity level are used for fabrication of superconducting radio frequency cavities (SCRF), and thermal conductivity is an important parameter influencing the performance of such SCRF cavities. We report here the temperature and magnetic field dependence of thermal conductivity {kappa} for superconducting niobium (Nb) samples, for which the electron mean free path I{sub e}, the phonon mean free path I{sub g}, and the vortex core diameter 2r{sub C} are of the same order of magnitude. The measured thermal conductivity is analyzed using the effective gap model (developed for I{sub e} >> 2r{sub C} (Dubeck et al 1963more » Phys. Rev. Lett. 10 98)) and the normal core model (developed for I{sub e} << 2r{sub C} (Ward and Dew-Hughes 1970 J. Phys. C: Solid St. Phys. 3 2245)). However, it is found that the effective gap model is not suitable for low temperatures when I{sub e} {approx} 2r{sub C}. The normal core model, on the other hand, is able to describe {kappa}(T,H) over the entire temperature range except in the field regime between H{sub C1} and H{sub C2} i.e. in the mixed state. It is shown that to understand the complete behavior of {kappa} in the mixed state, the scattering of quasi-particles from the vortex cores and the intervortex quasi-particle tunneling are to be invoked. The quasi-particle scattering from vortices for the present system is understood in terms of the framework of Sergeenkov and Ausloos (1995 Phys. Rev. B 52 3614) extending their approach to the case of Nb. The intervortex tunneling is understood within the framework of Schmidbauer et al (1970 Z. Phys. 240 30). Analysis of the field dependence of thermal conductivity shows that while the quasi-particle scattering from vortices dominates in the low fields, the intervortex quasi-particle tunneling dominates in high fields. Analysis of the temperature dependence of thermal conductivity shows that while the quasi-particle scattering is dominant at low temperatures, the intervortex quasi-particle tunneling is dominant at high temperatures.« less

Earth's Fiercely Cooling Core - 24 TW

NASA Astrophysics Data System (ADS)

Morgan, Jason P.; Vannucchi, Paola

2014-05-01

Earth's mantle and core are convecting planetary heat engines. The mantle convects to lose heat from slow cooling, internal radioactivity, and core heatflow across its base. Its convection generates plate tectonics, volcanism, and the loss of ~35 TW of mantle heat through Earth's surface. The core convects to lose heat from slow cooling, small amounts of internal radioactivity, and the freezing-induced growth of a compositionally denser inner core. Core convection produces the geodynamo generating Earth's geomagnetic field. The geodynamo was thought to be powered by ~4 TW of heatloss across the core-mantle boundary, a rate sustainable (cf. Gubbins et al., 2003; Nimmo, 2007) by freezing a compositionally denser inner core over the ~3 Ga that Earth is known to have had a strong geomagnetic field (cf. Tarduno, 2007). However, recent determinations of the outer core's thermal conductivity(Pozzo et al., 2012; Gomi et al., 2013) indicate that >15 TW of power should conduct down its adiabat. Conducted power is unavailable to drive thermal convection, implying that the geodynamo needs a long-lived >17 TW power source. Core cooling was thought too weak for this, based on estimates for the Clapeyron Slope for high-pressure freezing of an idealized pure-iron core. Here we show that the ~500-1000 kg/m3 seismically-inferred jump in density between the liquid outer core and solid inner core allows us to directly infer the core-freezing Clapeyron Slope for the outer core's actual composition which contains ~8±2% lighter elements (S,Si,O,Al, H,…) mixed into a Fe-Ni alloy. A PREM-like 600 kg/m3 - based Clapeyron Slope implies there has been ~774K of core cooling during the freezing and growth of the inner core, releasing ~24 TW of power during the past ~3 Ga. If so, core cooling can easily power Earth's long-lived geodynamo. Another major implication of ~24 TW heatflow across the core-mantle boundary is that the present-day mantle is strongly 'bottom-heated', and diapiric mantle plumes should dominate deep mantle upwelling.

High performance of SDC and GDC core shell type composite electrolytes using methane as a fuel for low temperature SOFC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Irshad, Muneeb; Siraj, Khurram, E-mail: razahussaini786@gmail.com, E-mail: khurram.uet@gmail.com; Javed, Fayyaz

Nanocomposites Samarium doped Ceria (SDC), Gadolinium doped Ceria (GDC), core shell SDC amorphous Na{sub 2}CO{sub 3} (SDCC) and GDC amorphous Na{sub 2}CO{sub 3} (GDCC) were synthesized using co-precipitation method and then compared to obtain better solid oxide electrolytes materials for low temperature Solid Oxide Fuel Cell (SOFCs). The comparison is done in terms of structure, crystallanity, thermal stability, conductivity and cell performance. In present work, XRD analysis confirmed proper doping of Sm and Gd in both single phase (SDC, GDC) and dual phase core shell (SDCC, GDCC) electrolyte materials. EDX analysis validated the presence of Sm and Gd in bothmore » single and dual phase electrolyte materials; also confirming the presence of amorphous Na{sub 2}CO{sub 3} in SDCC and GDCC. From TGA analysis a steep weight loss is observed in case of SDCC and GDCC when temperature rises above 725 °C while SDC and GDC do not show any loss. The ionic conductivity and cell performance of single phase SDC and GDC nanocomposite were compared with core shell GDC/amorphous Na{sub 2}CO{sub 3} and SDC/ amorphous Na{sub 2}CO{sub 3} nanocomposites using methane fuel. It is observed that dual phase core shell electrolytes materials (SDCC, GDCC) show better performance in low temperature range than their corresponding single phase electrolyte materials (SDC, GDC) with methane fuel.« less

Energy efficiency buildings program

NASA Astrophysics Data System (ADS)

1981-05-01

Progress is reported in developing techniques for auditing the energy performance of buildings. The ventilation of buildings and indoor air quality is discussed from the viewpoint of (1) combustion generated pollutants; (2) organic contaminants; (3) radon emanation, measurements, and control; (4) strategies for the field monitoring of indoor air quality; and (5) mechanical ventilation systems using air-to-air heat exchanges. The development of energy efficient windows to provide optimum daylight with minimal thermal losses in cold weather and minimum thermal gain in hot weather is considered as well as the production of high frequency solid state ballasts for fluorescent lights to provide more efficient lighting at a 25% savings over conventional core ballasts. Data compilation, analysis, and demonstration activities are summarized.

Preparation of Geophysical Fluid Flow Experiments With The Rotating Spherical Gap Flow Model In Space

NASA Astrophysics Data System (ADS)

Egbers, C.

The'GeoFlow' is an ESA experiment planned for the Fluid Science Laboratory on ISS under the scientific coordination (PI) of the Department of Aerodynamics and Fluid Mechanics (LAS) at the Brandenburg Technical University (BTU) of Cottbus, Germany. The objective of the experiment is to study thermal convection in the gap between two concentric rotating (full) spheres. A central symmetric force field simi- lar to the gravity field acting on planets can be produced by applying a high voltage between inner and outer sphere using the dielectrophoretic effect (rotating capacitor). To counter the unidirectional gravity under terrestrial conditions, this experiment re- quires a microgravity environment. The parameters of the experiment are chosen in analogy to the thermal convective motions in the outer core of the Earth. In analogy to geophysical motions in the Earth`s liquid core the experiment can rotate as solid body as well as differential (inner to outer). Thermal convection is produced by heat- ing the inner sphere and cooling the outer ones. Furtheron, the variation of radius ratio between inner and outer sphere is foreseen as a parameter variation. The flows to be investigated will strongly depend on the gap width and on the Prandtl number.

Design analysis and risk assessment for a single stage to orbit nuclear thermal rocket

NASA Astrophysics Data System (ADS)

Labib, Satira I.

Recent advances in high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This thesis describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1-15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 700 seconds. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. At the same power level, the 40 cm reactor results in the lowest radiation dose rate of the three reactors. Radiation dose rates decrease to background levels ~3.5 km from the launch site. After a one-year decay time, all of the activated materials produced by an NTR launch would be classified as Class A low-level waste. The activation of air produces significant amounts of argon-41 and nitrogen-16 within 100 m of the launch. The derived air concentration, DAC, from the activation products decays to less than unity within two days, with only argon-41 remaining. After 10 minutes of full power operation the 120 cm core corresponding to a 15 MT payload contains 2.5 x 1013, 1.4 x 1012, 1.5 x 1012, and 7.8 x 10 7 Bq of 131I, 137Cs, 90Sr, and 239Pu respectively. The decay heat after shutdown increases with increasing reactor power with a maximum decay heat of 108 kW immediately after shutdown for the 15 MT payload.

The nuclear battery

NASA Astrophysics Data System (ADS)

Kozier, K. S.; Rosinger, H. E.

The evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery is reviewed. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work.

Microchannel plate special nuclear materials sensor

NASA Astrophysics Data System (ADS)

Feller, W. B.; White, P. L.; White, P. B.; Siegmund, O. H. W.; Martin, A. P.; Vallerga, J. V.

2011-10-01

Nova Scientific Inc., is developing for the Domestic Nuclear Detection Office (DNDO SBIR #HSHQDC-08-C-00190), a solid-state, high-efficiency neutron detection alternative to 3He gas tubes, using neutron-sensitive microchannel plates (MCPs) containing 10B and/or Gd. This work directly supports DNDO development of technologies designed to detect and interdict nuclear weapons or illicit nuclear materials. Neutron-sensitized MCPs have been shown theoretically and more recently experimentally, to be capable of thermal neutron detection efficiencies equivalent to 3He gas tubes. Although typical solid-state neutron detectors typically have an intrinsic gamma sensitivity orders of magnitude higher than that of 3He gas detectors, we dramatically reduce gamma sensitivity by combining a novel electronic coincidence rejection scheme, employing a separate but enveloping gamma scintillator. This has already resulted in a measured gamma rejection ratio equal to a small 3He tube, without in principle sacrificing neutron detection efficiency. Ongoing improvements to the MCP performance as well as the coincidence counting geometry will be described. Repeated testing and validation with a 252Cf source has been underway throughout the Phase II SBIR program, with ongoing comparisons to a small commercial 3He gas tube. Finally, further component improvements and efforts toward integration maturity are underway, with the goal of establishing functional prototypes for SNM field testing.

In-situ Density and Thermal Expansion Measurements of Fe and Fe-S Alloying Liquids Under Planetary Core Conditions

NASA Astrophysics Data System (ADS)

Jing, Z.; Chantel, J.; Yu, T.; Sakamaki, T.; Wang, Y.

2015-12-01

Liquid iron is likely the dominant constituent in the cores of terrestrial planets and icy satellites such as Earth, Mars, Mercury, the Moon, Ganymede, and Io. Suggested by geophysical and geochemical observations, light elements such as S, C, Si, etc., are likely present in planetary cores. These light elements can significantly reduce the density and melting temperature of the Fe cores, and hence their abundances are crucial to our understanding of the structure and thermal history of planetary cores, as well as the generation of intrinsic magnetic fields. Knowledge on the density of Fe-light element alloying liquids at high pressures is critical to place constraints on the composition of planetary cores. However, density data on liquid Fe-light element alloys at core pressures are very limited in pressure and composition and are sometimes controversial. In this study, we extend the density dataset for Fe-rich liquids by measuring the density of Fe, Fe-10wt%S, Fe-20wt%S, Fe-27wt%S, and FeS liquids using the X-ray absorption technique in a DIA-type multianvil apparatus up to 7 GPa and 2173 K. An ion chamber (1D-detector) and a CCD camera (2D-detector) were used to measure intensities of transmitted monochromatic X-rays through molten samples, with the photon energy optimized at 40 keV. The densities were then determined from the Beer-Lambert law using the mass absorption coefficients, calibrated by solid standards using X-ray diffraction. At each pressure, density measurements were conducted at a range of temperatures above the liquidus of the samples, enabling the determination of thermal expansion. Combined with our previous results on the sound velocity of Fe and Fe-S liquids at high pressures (Jing et al., 2014, Earth Planet. Sci. Lett. 396, 78-87), these data provide tight constraints on the equation of state and thermodynamic properties such as the adiabatic temperature gradient for Fe-S liquids. We will discuss these results with implications to planetary cores.

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Proceedings of the AFOSR Special Conference on Prime-Power for High Energy Space Systems, Norfolk, Virginia, 22-25 February 1982. Volume I.

DTIC Science & Technology

1982-02-25

source both liquid and solid fuel combustion devices have been successfully demonstrated during various development programs . Nuclear reactor heat...U02 fuel in the core . Improving the heat pipe model to correlate more closely with the experimental data is a major concern in the development of...ORGANIZATION NAME AND ADDRESS 10. PROGRAM ELEMENT, PROJECT. TASK Research & Development Associates (RDA) AREA &WKNT AE (X Rosslyn, VA 22209 61102F 2301

Numerical simulation and experimental study of heat-fluid-solid coupling of double flapper-nozzle servo valve

NASA Astrophysics Data System (ADS)

Zhao, Jianhua; Zhou, Songlin; Lu, Xianghui; Gao, Dianrong

2015-09-01

The double flapper-nozzle servo valve is widely used to launch and guide the equipment. Due to the large instantaneous flow rate of servo valve working under specific operating conditions, the temperature of servo valve would reach 120°C and the valve core and valve sleeve deform in a short amount of time. So the control precision of servo valve significantly decreases and the clamping stagnation phenomenon of valve core appears. In order to solve the problem of degraded control accuracy and clamping stagnation of servo valve under large temperature difference circumstance, the numerical simulation of heat-fluid-solid coupling by using finite element method is done. The simulation result shows that zero position leakage of servo valve is basically impacted by oil temperature and change of fit clearance. The clamping stagnation is caused by warpage-deformation and fit clearance reduction of the valve core and valve sleeve. The distribution rules of the temperature and thermal-deformation of shell, valve core and valve sleeve and the pressure, velocity and temperature field of flow channel are also analyzed. Zero position leakage and electromagnet's current when valve core moves in full-stroke are tested using Electro-hydraulic Servo-valve Characteristic Test-bed of an aerospace sciences and technology corporation. The experimental results show that the change law of experimental current at different oil temperatures is roughly identical to simulation current. The current curve of the electromagnet is smooth when oil temperature is below 80°C, but the amplitude of current significantly increases and the hairy appears when oil temperature is above 80°C. The current becomes smooth again after the warped valve core and valve sleeve are reground. It indicates that clamping stagnation is caused by warpage-deformation and fit clearance reduction of valve core and valve sleeve. This paper simulates and tests the heat-fluid-solid coupling of double flapper-nozzle servo valve, and the obtained results provide the reference value for the design of double flapper-nozzle force feedback servo valve.

Stability Estimation of ABWR on the Basis of Noise Analysis

NASA Astrophysics Data System (ADS)

Furuya, Masahiro; Fukahori, Takanori; Mizokami, Shinya; Yokoya, Jun

In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel stability and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUS-F facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A real-time simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying auto regressive (AR) methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

Initial Coupling of the RELAP-7 and PRONGHORN Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi; D. Andrs; A.A. Bingham

2012-10-01

Modern nuclear reactor safety codes require the ability to solve detailed coupled neutronic- thermal fluids problems. For larger cores, this implies fully coupled higher dimensionality spatial dynamics with appropriate feedback models that can provide enough resolution to accurately compute core heat generation and removal during steady and unsteady conditions. The reactor analysis code PRONGHORN is being coupled to RELAP-7 as a first step to extend RELAP’s current capabilities. This report details the mathematical models, the type of coupling, and the testing results from the integrated system. RELAP-7 is a MOOSE-based application that solves the continuity, momentum, and energy equations inmore » 1-D for a compressible fluid. The pipe and joint capabilities enable it to model parts of the power conversion unit. The PRONGHORN application, also developed on the MOOSE infrastructure, solves the coupled equations that define the neutron diffusion, fluid flow, and heat transfer in a full core model. The two systems are loosely coupled to simplify the transition towards a more complex infrastructure. The integration is tested on a simplified version of the OECD/NEA MHTGR-350 Coupled Neutronics-Thermal Fluids benchmark model.« less

Low-Velocity Impact Behavior of Sandwich Structures with Additively Manufactured Polymer Lattice Cores

NASA Astrophysics Data System (ADS)

Turner, Andrew J.; Al Rifaie, Mohammed; Mian, Ahsan; Srinivasan, Raghavan

2018-05-01

Sandwich panel structures are widely used in aerospace, marine, and automotive applications because of their high flexural stiffness, strength-to-weight ratio, good vibration damping, and low through-thickness thermal conductivity. These structures consist of solid face sheets and low-density cellular core structures, which are traditionally based upon honeycomb folded-sheet topologies. The recent advances in additive manufacturing (AM) or 3D printing process allow lattice core configurations to be designed with improved mechanical properties. In this work, the sandwich core is comprised of lattice truss structures (LTS). Two different LTS designs are 3D-printed using acrylonitrile butadiene styrene (ABS) and are tested under low-velocity impact loads. The absorption energy and the failure mechanisms of lattice cells under such loads are investigated. The differences in energy-absorption capabilities are captured by integrating the load-displacement curve found from the impact response. It is observed that selective placement of vertical support struts in the unit-cell results in an increase in the absorption energy of the sandwich panels.

Low-Velocity Impact Behavior of Sandwich Structures with Additively Manufactured Polymer Lattice Cores

NASA Astrophysics Data System (ADS)

Turner, Andrew J.; Al Rifaie, Mohammed; Mian, Ahsan; Srinivasan, Raghavan

2018-04-01

Sandwich panel structures are widely used in aerospace, marine, and automotive applications because of their high flexural stiffness, strength-to-weight ratio, good vibration damping, and low through-thickness thermal conductivity. These structures consist of solid face sheets and low-density cellular core structures, which are traditionally based upon honeycomb folded-sheet topologies. The recent advances in additive manufacturing (AM) or 3D printing process allow lattice core configurations to be designed with improved mechanical properties. In this work, the sandwich core is comprised of lattice truss structures (LTS). Two different LTS designs are 3D-printed using acrylonitrile butadiene styrene (ABS) and are tested under low-velocity impact loads. The absorption energy and the failure mechanisms of lattice cells under such loads are investigated. The differences in energy-absorption capabilities are captured by integrating the load-displacement curve found from the impact response. It is observed that selective placement of vertical support struts in the unit-cell results in an increase in the absorption energy of the sandwich panels.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bsebsu, F.M.; Abotweirat, F.; Elwaer, S.

2008-07-15

The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulicmore » design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)« less

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forget, Benoit; Smith, Kord; Kumar, Shikhar

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevantmore » multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.« less

Bulk and Thin film Properties of Nanoparticle-based Ionic Materials

NASA Astrophysics Data System (ADS)

Fang, Jason

2008-03-01

Nanoparticle-based ionic materials (NIMS) offer exciting opportunities for research at the forefront of science and engineering. NIMS are hybrid particles comprised of a charged oligomeric corona attached to hard, inorganic nanoparticle cores. Because of their hybrid nature, physical properties --rheological, optical, electrical, thermal - of NIMS can be tailored over an unusually wide range by varying geometric and chemical characteristics of the core and canopy and thermodynamic variables such as temperature and volume fraction. On one end of the spectrum are materials with a high core content, which display properties similar to crystalline solids, stiff waxes, and gels. At the opposite extreme are systems that spontaneously form particle-based fluids characterized by transport properties remarkably similar to simple liquids. In this poster I will present our efforts to synthesize NIMS and discuss their bulk and surface properties. In particular I will discuss our work on preparing smart surfaces using NIMS.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Radiation doses to individuals due to ²³⁸U, ²³²Th and ²²²Rn from the immersion in thermal waters and to radon progeny from the inhalation of air inside thermal stations.

PubMed

Misdaq, M A; Ghilane, M; Ouguidi, J; Outeqablit, K

2012-11-01

In Morocco, thermal waters have been used for decades for the treatment of various diseases. To explore the exposure pathway of (238)U, (232)Th and (222)Rn to the skin of bathers from the immersion in thermal waters, these radionuclides were measured inside waters collected from different Moroccan thermal springs, by means of CR-39 and LR-115 type II solid-state nuclear track detectors (SSNTDs), and corresponding annual committed effective doses to skin were determined. Accordingly, to assess radiation dose due to radon short-lived decay products from the inhalation of air by individuals, concentrations of these radionuclides were measured in indoor air of two thermal stations by evaluating mean critical angles of etching of the CR-39 and LR-115 II SSNTDs. Committed effective doses due to the short-lived radon decay products (218)Po and (214)Po by bathers and working personnel inside the thermal stations studied were determined.

Discovery of massive star formation quenching by non-thermal effects in the centre of NGC 1097

NASA Astrophysics Data System (ADS)

Tabatabaei, F. S.; Minguez, P.; Prieto, M. A.; Fernández-Ontiveros, J. A.

2018-01-01

Observations show that massive star formation quenches first at the centres of galaxies. To understand quenching mechanisms, we investigate the thermal and non-thermal energy balance in the central kpc of NGC 1097—a prototypical galaxy undergoing quenching—and present a systematic study of the nuclear star formation efficiency and its dependencies. This region is dominated by the non-thermal pressure from the magnetic field, cosmic rays and turbulence. A comparison of the mass-to-magnetic flux ratio of the molecular clouds shows that most of them are magnetically critical or supported against the gravitational collapse needed to form the cores of massive stars. Moreover, the star formation efficiency of the clouds drops with the magnetic field strength. Such an anti-correlation holds with neither the turbulent nor the thermal pressure. Hence, a progressive build up of the magnetic field results in high-mass stars forming inefficiently, and this may be the cause of the low-mass stellar population in the bulges of galaxies.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Tidal deformations of compact stars with crystalline quark matter

NASA Astrophysics Data System (ADS)

Lau, S. Y.; Leung, P. T.; Lin, L.-M.

2017-05-01

We study the tidal deformability of bare quark stars and hybrid compact stars composed of a quark-matter core in general relativity, assuming that the deconfined quark matter exists in a crystalline color superconducting phase. We find that taking the elastic property of crystalline quark matter into account in the calculation of the tidal deformability can break the universal I-Love relation discovered for fluid compact stars, which connects the moment of inertia and tidal deformability. Our result suggests that measurements of the moment of inertia and tidal deformability can in principle be used to test the existence of solid quark stars, despite our ignorance of the high-density equation of state. Assuming that the moment of inertia can be measured to 10% level, one can then distinguish a 1.4 (1 ) M⊙ solid quark star described by our quark-matter equation of state model with a gap parameter Δ =25 MeV from a fluid compact star if the tidal deformability can be measured to about 10% (45%) level. On the other hand, we find that the nuclear matter fluid envelope of a hybrid star can screen out the effect of the solid core significantly so that the resulting I-Love relation for hybrid stars still agrees with the universal relation for fluid stars to about 1% level.

LIQUID METAL COMPOSITIONS CONTAINING URANIUM

DOEpatents

Teitel, R.J.

1959-04-21

Liquid metal compositions containing a solid uranium compound dispersed therein is described. Uranium combines with tin to form the intermetallic compound USn/sub 3/. It has been found that this compound may be incorporated into a liquid bath containing bismuth and lead-bismuth components, if a relatively small percentage of tin is also included in the bath. The composition has a low thermal neutron cross section which makes it suitable for use in a liquid metal fueled nuclear reactor.

Iron-rich Oxides at the Core-mantle Boundary

NASA Astrophysics Data System (ADS)

Wicks, J. K.; Jackson, J. M.; Sturhahn, W.; Bower, D. J.; Zhuravlev, K. K.; Prakapenka, V.

2013-12-01

Seismic observations near the base of the core-mantle boundary (CMB) have detected 5-20 km thick patches in which the seismic wave velocities are reduced by up to 30%. These ultra-low velocity zones (ULVZs) have been interpreted as aggregates of partially molten material (e.g. Williams and Garnero 1996, Hernlund and Jellinek, 2010) or as solid, iron-enriched residues (e.g. Knittle and Jeanloz, 1991; Mao et al., 2006; Wicks et al., 2010), typically based on proposed sources of velocity reduction. The stabilities of these structure types have been explored through dynamic models that have assembled a relationship between ULVZ stability and density (Hernlund and Tackley, 2007; Bower et al., 2010). Now, to constrain the chemistry and mineralogy of ULVZs, more information is needed on the relationship between density and sound velocity of candidate phases. We present the pressure-volume-temperature equation of state of (Mg0.06 57Fe0.94)O determined up to pressures of 120 GPa and temperatures of 2000 K. Volume was measured with X-ray diffraction at beamline 13-ID-D of the Advanced Photon Source (APS), where high pressures and temperatures are achieved in a diamond anvil cell with in-situ laser heating. Sample assemblies were prepared using dehydrated NaCl as an insulator and neon as a pressure transmitting medium. We present results with and without iron as a buffer and thermal pressure gauge. We have also determined the room temperature Debye velocity (VD) of (Mg0.06 57Fe0.94)O using nuclear resonant inelastic x-ray scattering and in-situ X-ray diffraction, up to 80 GPa at 3-ID-B of the APS. The effect of the electronic environment of the iron sites on the velocities was tracked in-situ using synchrotron Moessbauer spectroscopy. Using our measured equation of state, the seismically relevant compressional (VP) and shear (VS) wave velocities were calculated from the Debye velocities. We combine these studies with a simple mixing model to predict the properties of a solid ULVZ and show that a small amount of iron-rich (Mg,Fe)O can greatly reduce the average sound velocity of an aggregate assemblage. When combined with a geodynamic model of a solid ULVZ (Bower et al., 2011), we can directly correlate inferred sound velocities to mineralogy and predicted ULVZ shapes. In this presentation, our combined geodynamic and mineral physics model of a solid ULVZ will be used to explore the relationship between the observed sound velocities and mineralogy of ULVZs with added insight into ULVZ morphology.

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

NASA Astrophysics Data System (ADS)

Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

2018-01-01

In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

Engine System Model Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Nelson, Karl W.; Simpson, Steven P.

2006-01-01

In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

Atmospheric Mining in the Outer Solar System:. [Aerial Vehicle Reconnaissance and Exploration Options

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan A.

2014-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and hydrogen can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and hydrogen (deuterium, etc.) were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate, storage options, and different methods of direct use of the captured gases. Additional supporting analyses were conducted to illuminate vehicle sizing and orbital transportation issues. While capturing 3He, large amounts of hydrogen and 4He are produced. With these two additional gases, the potential for fueling small and large fleets of additional exploration and exploitation vehicles exists. Additional aerospacecraft or other aerial vehicles (UAVs, balloons, rockets, etc.) could fly through the outer planet atmospheres, for global weather observations, localized storm or other disturbance investigations, wind speed measurements, polar observations, etc. Deep-diving aircraft (built with the strength to withstand many atmospheres of pressure) powered by the excess hydrogen or helium 4 may be designed to probe the higher density regions of the gas giants. Outer planet atmospheric properties, atmospheric storm data, and mission planning for future outer planet UAVs are presented.

Atmospheric Mining in the Outer Solar System: Outer Planet Orbital Transfer and Lander Analyses

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2016-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and deuterium can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and deuterium were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate, storage options, and different methods of direct use of the captured gases. While capturing 3He, large amounts of hydrogen and 4He are produced. Analyses of orbital transfer vehicles (OTVs), landers, and the issues with in-situ resource utilization (ISRU) mining factories are included. Preliminary observations are presented on near-optimal selections of moon base orbital locations, OTV power levels, and OTV and lander rendezvous points. For analyses of round trip OTV flights from Uranus to Miranda or Titania, a 10-Megawatt electric (MWe) OTV power level and a 200-metric ton (MT) lander payload were selected based on a relative short OTV trip time and minimization of the number of lander flights. A similar optimum power level is suggested for OTVs flying from low orbit around Neptune to Thalassa or Triton. Several moon base sites at Uranus and Neptune and the OTV requirements to support them are also addressed.

Atmospheric Mining in the Outer Solar System: Outer Planet Orbital Transfer and Lander Analyses

NASA Technical Reports Server (NTRS)

Palaszewski, Bryan

2016-01-01

Atmospheric mining in the outer solar system has been investigated as a means of fuel production for high energy propulsion and power. Fusion fuels such as Helium 3 (3He) and deuterium can be wrested from the atmospheres of Uranus and Neptune and either returned to Earth or used in-situ for energy production. Helium 3 and deuterium were the primary gases of interest with hydrogen being the primary propellant for nuclear thermal solid core and gas core rocket-based atmospheric flight. A series of analyses were undertaken to investigate resource capturing aspects of atmospheric mining in the outer solar system. This included the gas capturing rate, storage options, and different methods of direct use of the captured gases. While capturing 3He, large amounts of hydrogen and 4He are produced. Analyses of orbital transfer vehicles (OTVs), landers, and the issues with in-situ resource utilization (ISRU) mining factories are included. Preliminary observations are presented on near-optimal selections of moon base orbital locations, OTV power levels, and OTV and lander rendezvous points. For analyses of round trip OTV flights from Uranus to Miranda or Titania, a 10- Megawatt electric (MWe) OTV power level and a 200 metricton (MT) lander payload were selected based on a relative short OTV trip time and minimization of the number of lander flights. A similar optimum power level is suggested for OTVs flying from low orbit around Neptune to Thalassa or Triton. Several moon base sites at Uranus and Neptune and the OTV requirements to support them are also addressed.

Ab Initio Investigations of High-Pressure Melting of Dense Lithium

NASA Astrophysics Data System (ADS)

Clay, Raymond; Morales, Miguel; Bonev, Stanimir

Lithium at ambient conditions is the simplest alkali metal and exhibits textbook nearly-free electron behavior. As the density is increased, however, significant core/valence overlap leads to surprisingly complex chemistry. We have systematically investigated the phase diagram of lithium at pressures ranging between two and six million atmospheres. Through a combination of density functional theory based path-integral and classical molecular dynamics simulations, we have investigated the impact of both nuclear quantum effects and anharmonicity on the melting line and solid phase boundaries. We also investigate how the inclusion of nuclear quantum effects and approximations in the treatment of electronic exchange-correlation impact the robustness of previous predictions of tetrahedral clustering in dense liquid Li. Sandia National Laboratories is a multi-mission laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT).

PubMed

Payne, Liam; Heard, Peter J; Scott, Thomas B

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK's first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600-1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment.

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT)

PubMed Central

Payne, Liam; Heard, Peter J.; Scott, Thomas B.

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK’s first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600–1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment. PMID:26575374

Nuclear Cryogenic Propulsion Stage Conceptual Design and Mission Analysis

NASA Technical Reports Server (NTRS)

Kos, Larry D.; Russell, Tiffany E.

2014-01-01

The Nuclear Cryogenic Propulsion Stage (NCPS) is an in-space transportation vehicle, comprised of three main elements, designed to support a long-stay human Mars mission architecture beginning in 2035. The stage conceptual design and the mission analysis discussed here support the current nuclear thermal propulsion going on within partnership activity of NASA and the Department of Energy (DOE). The transportation system consists of three elements: 1) the Core Stage, 2) the In-line Tank, and 3) the Drop Tank. The driving mission case is the piloted flight to Mars in 2037 and will be the main point design shown and discussed. The corresponding Space Launch System (SLS) launch vehicle (LV) is also presented due to it being a very critical aspect of the NCPS Human Mars Mission architecture due to the strong relationship between LV lift capability and LV volume capacity.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

Comparative Analysis of Hexagonal Solid Silica and Nitro-benzene Filled Hollow Core Photonic Crystal Fiber

NASA Astrophysics Data System (ADS)

Shahiruddin; Singh, Dharmendra K.; Hassan, M. A.

2018-02-01

A comparative study of five ring solid core and nitrobenzene filled hollow core liquid filled photonic crystal fiber (PCF) are presented. Considering the same structure, one is used as solid silica and another one is filled with nitrobenzene in the core. Here the paper elaborates the confinement loss, dispersion properties and birefringence of an index-guiding PCF with asymmetric cladding designed and analyzed by the finite-element method. The proposed structure shows the low confinement loss in case of solid silica, negative dispersion in nitrobenzene filled hollow core PCF and high birefringence in both the cases. The calculated values shows flat zero confinement loss in 0.7 µm to 1.54 µm range, flat zero dispersion is achieved in solid core and -2000 ps/km-nm in nitrobenzene filled hollow core PCF and high birefringence in the range of 10-3 in nitrobenzene filled hollow core PCF. Results show the relative analysis at different air fill fraction.

Constraints on The Coupled Thermal Evolution of the Earth's Core and Mantle, The Age of The Inner Core, And The Origin of the 186Os/188Os Core(?) Signal in Plume-Derived Lavas

NASA Astrophysics Data System (ADS)

Lassiter, J. C.

2005-12-01

Thermal and chemical interaction between the core and mantle has played a critical role in the thermal and chemical evolution of the Earth's interior. Outer core convection is driven by core cooling and inner core crystallization. Core/mantle heat transfer also buffers mantle potential temperature, resulting in slower rates of mantle cooling (~50-100 K/Ga) than would be predicted from the discrepancy between current rates of surface heat loss (~44 TW) and internal radioactive heat production (~20 TW). Core/mantle heat transfer may also generate thermal mantle plumes responsible for ocean island volcanic chains such as the Hawaiian Islands. Several studies suggest that mantle plumes, in addition to transporting heat from the core/mantle boundary, also carry a chemical signature of core/mantle interaction. Elevated 186Os/188Os ratios in lavas from Hawaii, Gorgona, and in the 2.8 Ga Kostomuksha komatiites have been interpreted as reflecting incorporation of an outer core component with high time-integrated Pt/Os and Re/Os ( Brandon et al., 1999, 2003; Puchtel et al., 2005). Preferential partitioning of Os relative to Re and Pt into the inner core during inner core growth may generate elevated Re/Os and Pt/Os ratios in the residual outer core. Because of the long half-life of 190Pt (the parent of 186Os, t1/2 = 489 Ga), an elevated 186Os/188Os outer core signature in plume lavas requires that inner core crystallization began early in Earth history, most likely prior to 3.5 Ga. This in turn requires low time-averaged core/mantle heat flow (<~2.5 TW) or large quantities of heat-producing elements in the core. Core/mantle heat flow may be estimated using boundary-layer theory, by measuring the heat transported in mantle plumes, by estimating the heat transported along the outer core adiabat, or by comparing the rates of heat production, surface heat loss, and secular cooling of the mantle. All of these independent methods suggest time-averaged core/mantle heat flow of ~5-14 TW. In the absence of heat-producing elements in the core, such high heat flow rates require an inner core younger than ~1 Ga and preclude the development of significant 186Os enrichment in the outer core. Experimental studies suggest that potassium may partition into Fe-S-O liquids during core formation. Radioactive decay of potassium in the core could provide an additional heat source and reconcile geophysical evidence for high core/mantle heat flow with apparent geochemical evidence for an ancient inner core. However, high concentrations of chalcophile elements such as Cu in the mantle are inconsistent with significant segregation of a S-rich liquid during core formation, precluding K partitioning into the core by this mechanism. Furthermore, core formation scenarios that would lead to high K content in the core (e.g., core formation prior to terrestrial volatile depletion) also result in high core Pb concentrations. Core/mantle interaction would then produce strong negative correlations between 186Os/188Os and 207Pb/204Pb ratios, but such correlations are not observed. In summary, elevated 186Os/188Os ratios in some plume-derived lavas are unlikely to reflect core/mantle interaction because the inner core is too young for this isotopic signature to have developed in the outer core. Melt generation from pyroxenite or fractionation of PGEs between sulfide melts and monosulfide solid solutions provide alternative mechanisms for generating ancient mantle reservoirs with elevated Pt/Os and 186Os/188Os.

First-principles analysis of anharmonic nuclear motion and thermal transport in thermoelectric materials

NASA Astrophysics Data System (ADS)

Tadano, Terumasa; Tsuneyuki, Shinji

2015-12-01

We show a first-principles approach for analyzing anharmonic properties of lattice vibrations in solids. We firstly extract harmonic and anharmonic force constants from accurate first-principles calculations based on the density functional theory. Using the many-body perturbation theory of phonons, we then estimate the phonon scattering probability due to anharmonic phonon-phonon interactions. We show the validity of the approach by computing the lattice thermal conductivity of Si, a typical covalent semiconductor, and selected thermoelectric materials PbTe and Bi2Te3 based on the Boltzmann transport equation. We also show that the phonon lifetime and the lattice thermal conductivity of the high-temperature phase of SrTiO3 can be estimated by employing the perturbation theory on top of the solution of the self-consistent phonon equation.

Depletion optimization of lumped burnable poisons in pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kodah, Z.H.

1982-01-01

Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solidmore » cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration.« less

Nuclear reactor apparatus

DOEpatents

Wade, Elman E.

1978-01-01

A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

Mechanical, Thermal and Acoustic Properties of Open-pore Phenolic Multi-structured Cryogel

NASA Astrophysics Data System (ADS)

Yao, Rui; Yao, Zhengjun; Zhou, Jintang; Liu, Peijiang; Lei, Yiming

2017-09-01

Open-pore phenolic cryogel acoustic multi-structured plates (OCMPs) were prepared via modified sol gel polymerization and freeze-dried methods. The pore morphology, mechanical, thermal and acoustic properties of the cryogels were investigated. From the experimental results, the cryogels exhibited a porous sandwich microstructure: A nano-micron double-pore structure was observed in the core layer of the plates, and nanosized pores were observed in the inner part of the micron pores. In addtion, compared with cryogel plates with uniform-pore (OCPs), the OCMPs had lower thermal conductivities. What’s more, the compressive and tensile strength of the OCMPs were much higher than those of OCPs. Finally, the OCMPs exhibited superior acoustic performances (20% solid content OCMPs performed the best) as compared with those of OCPs. Moreover, the sound insulation value and sound absorption bandwidth of OCMPs exhibited an improvement of approximately 3 and 2 times as compared with those of OCPs, respectively.

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

NASA Astrophysics Data System (ADS)

Kochetkov, Anatoly; Krása, Antonín; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente; Bianchini, Giancarlo; Fabrizio, Valentina; Carta, Mario; Firpo, Gabriele; Fridman, Emil; Sarotto, Massimo

2017-09-01

During the GUINEVERE FP6 European project (2006-2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

Observations on the Chernobyl Disaster and LNT

PubMed Central

Jaworowski, Zbigniew

2010-01-01

The Chernobyl accident was probably the worst possible catastrophe of a nuclear power station. It was the only such catastrophe since the advent of nuclear power 55 years ago. It resulted in a total meltdown of the reactor core, a vast emission of radionuclides, and early deaths of only 31 persons. Its enormous political, economic, social and psychological impact was mainly due to deeply rooted fear of radiation induced by the linear non-threshold hypothesis (LNT) assumption. It was a historic event that provided invaluable lessons for nuclear industry and risk philosophy. One of them is demonstration that counted per electricity units produced, early Chernobyl fatalities amounted to 0.86 death/GWe-year), and they were 47 times lower than from hydroelectric stations (∼40 deaths/GWe-year). The accident demonstrated that using the LNT assumption as a basis for protection measures and radiation dose limitations was counterproductive, and lead to sufferings and pauperization of millions of inhabitants of contaminated areas. The projections of thousands of late cancer deaths based on LNT, are in conflict with observations that in comparison with general population of Russia, a 15% to 30% deficit of solid cancer mortality was found among the Russian emergency workers, and a 5% deficit solid cancer incidence among the population of most contaminated areas. PMID:20585443

Observations on the Chernobyl Disaster and LNT.

PubMed

Jaworowski, Zbigniew

2010-01-28

The Chernobyl accident was probably the worst possible catastrophe of a nuclear power station. It was the only such catastrophe since the advent of nuclear power 55 years ago. It resulted in a total meltdown of the reactor core, a vast emission of radionuclides, and early deaths of only 31 persons. Its enormous political, economic, social and psychological impact was mainly due to deeply rooted fear of radiation induced by the linear non-threshold hypothesis (LNT) assumption. It was a historic event that provided invaluable lessons for nuclear industry and risk philosophy. One of them is demonstration that counted per electricity units produced, early Chernobyl fatalities amounted to 0.86 death/GWe-year), and they were 47 times lower than from hydroelectric stations ( approximately 40 deaths/GWe-year). The accident demonstrated that using the LNT assumption as a basis for protection measures and radiation dose limitations was counterproductive, and lead to sufferings and pauperization of millions of inhabitants of contaminated areas. The projections of thousands of late cancer deaths based on LNT, are in conflict with observations that in comparison with general population of Russia, a 15% to 30% deficit of solid cancer mortality was found among the Russian emergency workers, and a 5% deficit solid cancer incidence among the population of most contaminated areas.

The Last Minutes of Oxygen Shell Burning in a Massive Star

NASA Astrophysics Data System (ADS)

Müller, Bernhard; Viallet, Maxime; Heger, Alexander; Janka, Hans-Thomas

2016-12-01

We present the first 4π-three-dimensional (3D) simulation of the last minutes of oxygen shell burning in an 18 M ⊙ supernova progenitor up to the onset of core collapse. A moving inner boundary is used to accurately model the contraction of the silicon and iron core according to a one-dimensional stellar evolution model with a self-consistent treatment of core deleptonization and nuclear quasi-equilibrium. The simulation covers the full solid angle to allow the emergence of large-scale convective modes. Due to core contraction and the concomitant acceleration of nuclear burning, the convective Mach number increases to ˜0.1 at collapse, and an ℓ = 2 mode emerges shortly before the end of the simulation. Aside from a growth of the oxygen shell from 0.51 M ⊙ to 0.56 M ⊙ due to entrainment from the carbon shell, the convective flow is reasonably well described by mixing-length theory, and the dominant scales are compatible with estimates from linear stability analysis. We deduce that artificial changes in the physics, such as accelerated core contraction, can have precarious consequences for the state of convection at collapse. We argue that scaling laws for the convective velocities and eddy sizes furnish good estimates for the state of shell convection at collapse and develop a simple analytic theory for the impact of convective seed perturbations on shock revival in the ensuing supernova. We predict a reduction of the critical luminosity for explosion by 12%-24% due to seed asphericities for our 3D progenitor model relative to the case without large seed perturbations.

Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark; Gleicher, Frederick; Ortensi, Javier

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in amore » graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.« less

VB12-coated Gel-Core-SLN containing insulin: Another way to improve oral absorption.

PubMed

He, Haibing; Wang, Puxiu; Cai, Cuifang; Yang, Rui; Tang, Xing

2015-09-30

To improve the oral absorption of insulin, a novel carrier of Vitamin B12 (VB12) gel core solid lipid nanopaticles (Gel-Core-SLN, GCSLN) was designed with a gel core, lipid matrix and VB12-coated surface. VB12-stearate was synthesized and characterized by infrared spectroscopy (IR), nuclear magnetic resonance spectroscopy (NMR) and mass spectrometry (MS). Sol-gel conversion following ultrasonic heating and double emulsion technology were combined to implant the insulin-containing gel into solid lipid nanoparticles (SLN). The influence of the mode of administration, food, the amount of VB12-stearate and the particle size on the oral absorption of insulin incorporated in the VB12-GCSLN was investigated. The determined partition coefficient (LogP) of VB12-stearate in a dichloromethane (DCM)-water system was 3.4. This new structure of VB12-GCSLN had higher insulin encapsulation efficiency (EE) of 55.9%, a lower burst release of less than 10% in the first 2h. In vivo studies demonstrated that stronger absorption of insulin with a relative pharmacological availability (PA) of 9.31% compared with the normal insulin-loaded SLN and GCSLN and fairly stable blood glucose levels up to 12h were maintained without any sharp fluctuations. This study suggests that VB12-GCSLN containing insulin appears to be a promising nano carrier for oral delivery of biomacromolecules with relatively high pharmacological availability. Copyright © 2015 Elsevier B.V. All rights reserved.

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance duringmore » LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.« less

High-Pressure Geophysical Properties of Fcc Phase FeHX

NASA Astrophysics Data System (ADS)

Thompson, E. C.; Davis, A. H.; Bi, W.; Zhao, J.; Alp, E. E.; Zhang, D.; Greenberg, E.; Prakapenka, V. B.; Campbell, A. J.

2018-01-01

Face centered cubic (fcc) FeHX was synthesized at pressures of 18-68 GPa and temperatures exceeding 1,500 K. Thermally quenched samples were evaluated using synchrotron X-ray diffraction (XRD) and nuclear resonant inelastic X-ray scattering (NRIXS) to determine sample composition and sound velocities to 82 GPa. To aid in the interpretation of nonideal (X ≠ 1) stoichiometries, two equations of state for fcc FeHX were developed, combining an empirical equation of state for iron with two distinct synthetic compression curves for interstitial hydrogen. Matching the density deficit of the Earth's core using these equations of state requires 0.8-1.1 wt % hydrogen at the core-mantle boundary and 0.2-0.3 wt % hydrogen at the interface of the inner and outer cores. Furthermore, a comparison of Preliminary Reference Earth Model (PREM) to a Birch's law extrapolation of our experimental results suggests that an iron alloy containing ˜0.8-1.3 wt % hydrogen could reproduce both the density and compressional velocity (VP) of the Earth's outer core.

77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-23

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

Optical Sensors for Monitoring Gamma and Neutron Radiation

NASA Technical Reports Server (NTRS)

Boyd, Clark D.

2011-01-01

For safety and efficiency, nuclear reactors must be carefully monitored to provide feedback that enables the fission rate to be held at a constant target level via adjustments in the position of neutron-absorbing rods and moderating coolant flow rates. For automated reactor control, the monitoring system should provide calibrated analog or digital output. The sensors must survive and produce reliable output with minimal drift for at least one to two years, for replacement only during refueling. Small sensor size is preferred to enable more sensors to be placed in the core for more detailed characterization of the local fission rate and fuel consumption, since local deviations from the norm tend to amplify themselves. Currently, reactors are monitored by local power range meters (LPRMs) based on the neutron flux or gamma thermometers based on the gamma flux. LPRMs tend to be bulky, while gamma thermometers are subject to unwanted drift. Both electronic reactor sensors are plagued by electrical noise induced by ionizing radiation near the reactor core. A fiber optic sensor system was developed that is capable of tracking thermal neutron fluence and gamma flux in order to monitor nuclear reactor fission rates. The system provides near-real-time feedback from small- profile probes that are not sensitive to electromagnetic noise. The key novel feature is the practical design of fiber optic radiation sensors. The use of an actinoid element to monitor neutron flux in fiber optic EFPI (extrinsic Fabry-Perot interferometric) sensors is a new use of material. The materials and structure used in the sensor construction can be adjusted to result in a sensor that is sensitive to just thermal, gamma, or neutron stimulus, or any combination of the three. The tested design showed low sensitivity to thermal and gamma stimuli and high sensitivity to neutrons, with a fast response time.

Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

NASA Astrophysics Data System (ADS)

Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

Time-Dependent Simulations of the Formation and Evolution of Disk-Accreted Atmospheres Around Terrestrial Planets

NASA Astrophysics Data System (ADS)

Stoekl, Alexander; Dorfi, Ernst

2014-05-01

In the early, embedded phase of evolution of terrestrial planets, the planetary core accumulates gas from the circumstellar disk into a planetary envelope. This atmosphere is very significant for the further thermal evolution of the planet by forming an insulation around the rocky core. The disk-captured envelope is also the staring point for the atmospheric evolution where the atmosphere is modified by outgassing from the planetary core and atmospheric mass loss once the planet is exposed to the radiation field of the host star. The final amount of persistent atmosphere around the evolved planet very much characterizes the planet and is a key criterion for habitability. The established way to study disk accumulated atmospheres are hydrostatic models, even though in many cases the assumption of stationarity is unlikely to be fulfilled. We present, for the first time, time-dependent radiation hydrodynamics simulations of the accumulation process and the interaction between the disk-nebula gas and the planetary core. The calculations were performed with the TAPIR-Code (short for The adaptive, implicit RHD-Code) in spherical symmetry solving the equations of hydrodynamics, gray radiative transport, and convective energy transport. The models range from the surface of the solid core up to the Hill radius where the planetary envelope merges into the surrounding protoplanetary disk. Our results show that the time-scale of gas capturing and atmospheric growth strongly depends on the mass of the solid core. The amount of atmosphere accumulated during the lifetime of the protoplanetary disk (typically a few Myr) varies accordingly with the mass of the planet. Thus, a core with Mars-mass will end up with about 10 bar of atmosphere while for an Earth-mass core, the surface pressure reaches several 1000 bar. Even larger planets with several Earth masses quickly capture massive envelopes which in turn become gravitationally unstable leading to runaway accretion and the eventual formation of a gas planet.

Safety monitoring and reactor transient interpreter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hench, J. E.; Fukushima, T. Y.

1983-12-20

An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18,20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24)more » for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18,20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.« less

Flow instability in particle-bed nuclear reactors

NASA Astrophysics Data System (ADS)

Kerrebrock, Jack L.

The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded that operation at design exit temperature but at reduced power could be hazardous for such a reactor. But the calculations also show that an appropriate cold frit could very likely cure the instability. More definite conclusions must await calculations for specific designs.

Flow instability in particle-bed nuclear reactors

NASA Technical Reports Server (NTRS)

Kerrebrock, Jack L.

1993-01-01

The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded that operation at design exit temperature but at reduced power could be hazardous for such a reactor. But the calculations also show that an appropriate cold frit could very likely cure the instability. More definite conclusions must await calculations for specific designs.

Nanoscale phase engineering of thermal transport with a Josephson heat modulator.

PubMed

Fornieri, Antonio; Blanc, Christophe; Bosisio, Riccardo; D'Ambrosio, Sophie; Giazotto, Francesco

2016-03-01

Macroscopic quantum phase coherence has one of its pivotal expressions in the Josephson effect, which manifests itself both in charge and energy transport. The ability to master the amount of heat transferred through two tunnel-coupled superconductors by tuning their phase difference is the core of coherent caloritronics, and is expected to be a key tool in a number of nanoscience fields, including solid-state cooling, thermal isolation, radiation detection, quantum information and thermal logic. Here, we show the realization of the first balanced Josephson heat modulator designed to offer full control at the nanoscale over the phase-coherent component of thermal currents. Our device provides magnetic-flux-dependent temperature modulations up to 40 mK in amplitude with a maximum of the flux-to-temperature transfer coefficient reaching 200 mK per flux quantum at a bath temperature of 25 mK. Foremost, it demonstrates the exact correspondence in the phase engineering of charge and heat currents, breaking ground for advanced caloritronic nanodevices such as thermal splitters, heat pumps and time-dependent electronic engines.

Nanoscale phase engineering of thermal transport with a Josephson heat modulator

NASA Astrophysics Data System (ADS)

Fornieri, Antonio; Blanc, Christophe; Bosisio, Riccardo; D'Ambrosio, Sophie; Giazotto, Francesco

2016-03-01

Macroscopic quantum phase coherence has one of its pivotal expressions in the Josephson effect, which manifests itself both in charge and energy transport. The ability to master the amount of heat transferred through two tunnel-coupled superconductors by tuning their phase difference is the core of coherent caloritronics, and is expected to be a key tool in a number of nanoscience fields, including solid-state cooling, thermal isolation, radiation detection, quantum information and thermal logic. Here, we show the realization of the first balanced Josephson heat modulator designed to offer full control at the nanoscale over the phase-coherent component of thermal currents. Our device provides magnetic-flux-dependent temperature modulations up to 40 mK in amplitude with a maximum of the flux-to-temperature transfer coefficient reaching 200 mK per flux quantum at a bath temperature of 25 mK. Foremost, it demonstrates the exact correspondence in the phase engineering of charge and heat currents, breaking ground for advanced caloritronic nanodevices such as thermal splitters, heat pumps and time-dependent electronic engines.

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

NASA Technical Reports Server (NTRS)

Sorensen, Kirk; Juhasz, Albert

2007-01-01

Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

Determination of the core temperature of a Li-ion cell during thermal runaway

NASA Astrophysics Data System (ADS)

Parhizi, M.; Ahmed, M. B.; Jain, A.

2017-12-01

Safety and performance of Li-ion cells is severely affected by thermal runaway where exothermic processes within the cell cause uncontrolled temperature rise, eventually leading to catastrophic failure. Most past experimental papers on thermal runaway only report surface temperature measurement, while the core temperature of the cell remains largely unknown. This paper presents an experimentally validated method based on thermal conduction analysis to determine the core temperature of a Li-ion cell during thermal runaway using surface temperature and chemical kinetics data. Experiments conducted on a thermal test cell show that core temperature computed using this method is in good agreement with independent thermocouple-based measurements in a wide range of experimental conditions. The validated method is used to predict core temperature as a function of time for several previously reported thermal runaway tests. In each case, the predicted peak core temperature is found to be several hundreds of degrees Celsius higher than the measured surface temperature. This shows that surface temperature alone is not sufficient for thermally characterizing the cell during thermal runaway. Besides providing key insights into the fundamental nature of thermal runaway, the ability to determine the core temperature shown here may lead to practical tools for characterizing and mitigating thermal runaway.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nekoogar, F; Dowla, F; Wang, T

Recent advancements in the ultra-wide band Radio Frequency Identification (RFID) technology and solid state pillar type neutron detectors have enabled us to move forward in combining both technologies for advanced neutron monitoring. The LLNL RFID tag is totally passive and will operate indefinitely without the need for batteries. The tag is compact, can be directly mounted on metal, and has high performance in dense and cluttered environments. The LLNL coin-sized pillar solid state neutron detector has achieved a thermal neutron detection efficiency of 20% and neutron/gamma discrimination of 1E5. These performance values are comparable to a fieldable {sup 3}He basedmore » detector. In this paper we will discuss features about the two technologies and some potential applications for the advanced safeguarding of nuclear materials.« less

Design and calculation of low infrared transmittance and low emissivity coatings for heat radiative applications

NASA Astrophysics Data System (ADS)

Wang, Guang-Hai; Zhang, Yue; Zhang, Da-Hai; Fan, Jin-Peng

2012-02-01

The infrared transmittance and emissivity of heat-insulating coatings pigmented with various structural particles were studied using Kubelka-Munk theory and Mie theory. The primary design purpose was to obtain the low transmittance and low emissivity coatings to reduce the heat transfer by thermal radiation for high-temperature applications. In the case of silica coating layers constituted with various structural titania particles (solid, hollow, and core-shell spherical), the dependence of transmittance and emissivity of the coating layer on the particle structure and the layer thickness was investigated and optimized. The results indicate that the coating pigmented with core-shell titania particles exhibits a lower infrared transmittance and a lower emissivity value than that with other structural particles and is suitable to radiative heat-insulating applications.

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2005-01-01

The EFF-TF provides a facility to experimentally evaluate thermal hydraulic issues through the use of highly effective non-nuclear testing. These techniques provide a rapid, more cost effective method of evaluating designs and support development risk mitigation when concerns are associated with non-nuclear aspects of space nuclear systems. For many systems, electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004. Initial evaluation of the SAFE-100a (19 module stainless steel/sodium heat pipe reactor with integral gas neat exchanger) was performed with tests up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium SAFE-100 heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37-fuel pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to field a near term space nuclear system. Efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

Imaging the Moon's Core with Seismology

NASA Technical Reports Server (NTRS)

Weber, Renee C.; Lin, Pei-Ying Patty; Garnero, Ed J.; Williams, Quetin C.; Lognonne, Philippe

2011-01-01

Constraining the structure of the lunar core is necessary to improve our understanding of the present-day thermal structure of the interior and the history of a lunar dynamo, as well as the origin and thermal and compositional evolution of the Moon. We analyze Apollo deep moonquake seismograms using terrestrial array processing methods to search for the presence of reflected and converted energy from the lunar core. Although moonquake fault parameters are not constrained, we first explore a suite of theoretical focal spheres to verify that fault planes exist that can produce favorable core reflection amplitudes relative to direct up-going energy at the Apollo stations. Beginning with stacks of event seismograms from the known distribution of deep moonquake clusters, we apply a polarization filter to account for the effects of seismic scattering that (a) partitions energy away from expected components of ground motion, and (b) obscures all but the main P- and S-wave arrivals. The filtered traces are then shifted to the predicted arrival time of a core phase (e.g. PcP) and stacked to enhance subtle arrivals associated with the Moon s core. This combination of filtering and array processing is well suited for detecting deep lunar seismic reflections, since we do not expect scattered wave energy from near surface (or deeper) structure recorded at varying epicentral distances and stations from varying moonquakes at varying depths to stack coherently. Our results indicate the presence of a solid inner and fluid outer core, overlain by a partial-melt-containing boundary layer (Table 1). These layers are consistently observed among stacks from four classes of reflections: P-to-P, S-to-P, P-to-S, and S-to-S, and are consistent with current indirect geophysical estimates of core and deep mantle properties, including mass, moment of inertia, lunar laser ranging, and electromagnetic induction. Future refinements are expected following the successful launch of the GRAIL lunar orbiter and SELENE 2 lunar lander missions.

Quantum mechanical theory of dynamic nuclear polarization in solid dielectrics.

PubMed

Hu, Kan-Nian; Debelouchina, Galia T; Smith, Albert A; Griffin, Robert G

2011-03-28

Microwave driven dynamic nuclear polarization (DNP) is a process in which the large polarization present in an electron spin reservoir is transferred to nuclei, thereby enhancing NMR signal intensities. In solid dielectrics there are three mechanisms that mediate this transfer--the solid effect (SE), the cross effect (CE), and thermal mixing (TM). Historically these mechanisms have been discussed theoretically using thermodynamic parameters and average spin interactions. However, the SE and the CE can also be modeled quantum mechanically with a system consisting of a small number of spins and the results provide a foundation for the calculations involving TM. In the case of the SE, a single electron-nuclear spin pair is sufficient to explain the polarization mechanism, while the CE requires participation of two electrons and a nuclear spin, and can be used to understand the improved DNP enhancements observed using biradical polarizing agents. Calculations establish the relations among the electron paramagnetic resonance (EPR) and nuclear magnetic resonance (NMR) frequencies and the microwave irradiation frequency that must be satisfied for polarization transfer via the SE or the CE. In particular, if δ, Δ < ω(0I), where δ and Δ are the homogeneous linewidth and inhomogeneous breadth of the EPR spectrum, respectively, we verify that the SE occurs when ω(M) = ω(0S) ± ω(0I), where ω(M), ω(0S) and ω(0I) are, respectively, the microwave, and the EPR and NMR frequencies. Alternatively, when Δ > ω(0I) > δ, the CE dominates the polarization transfer. This two-electron process is optimized when ω(0S(1))-ω(0S(2)) = ω(0I) and ω(M)~ω(0S(1)) or ω(0S(2)), where ω(0S(1)) and ω(0S(2)) are the EPR Larmor frequencies of the two electrons. Using these matching conditions, we calculate the evolution of the density operator from electron Zeeman order to nuclear Zeeman order for both the SE and the CE. The results provide insights into the influence of the microwave irradiation field, the external magnetic field, and the electron-electron and electron-nuclear interactions on DNP enhancements.

On the Composition and Temperature of the Terrestrial Planetary Core

NASA Astrophysics Data System (ADS)

Fei, Yingwei

2013-06-01

The existence of liquid cores of terrestrial planets such as the Earth, Mar, and Mercury has been supported by various observation. The liquid state of the core provides a unique opportunity for us to estimate the temperature of the core if we know the melting temperature of the core materials at core pressure. Dynamic compression by shock wave, laser-heating in diamond-anvil cell, and resistance-heating in the multi-anvil device can melt core materials over a wide pressure range. There have been significant advances in both dynamic and static experimental techniques and characterization tool. In this tal, I will review some of the recent advances and results relevant to the composition and thermal state of the terrestrial core. I will also present new development to analyze the quenched samples recovered from laser-heating diamond-anvil cell experiments using combination of focused ion beam milling, high-resolution SEM imaging, and quantitative chemical analysi. With precision milling of the laser-heating spo, the melting point and element partitioning between solid and liquid can be precisely determined. It is also possible to re-construct 3D image of the laser-heating spot at multi-megabar pressures to better constrain melting point and understanding melting process. The new techniques allow us to extend precise measurements of melting relations to core pressures, providing better constraint on the temperature of the cor. The research is supported by NASA and NSF grants.

Advanced propulsion engine assessment based on a cermet reactor

NASA Technical Reports Server (NTRS)

Parsley, Randy C.

1993-01-01

A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

MEANS FOR CONTROLLING A NUCLEAR REACTOR

DOEpatents

Wilson, V.C.; Overbeck, W.P.; Slotin, L.; Froman, D.K.

1957-12-17

This patent relates to nuclear reactors of the type using a solid neutron absorbing material as a means for controlling the reproduction ratio of the system and thereby the power output. Elongated rods of neutron absorbing material, such as boron steel for example, are adapted to be inserted and removed from the core of tae reactor by electronic motors and suitable drive means. The motors and drive means are controlled by means responsive to the neutron density, such as ionization chambers. The control system is designed to be responsive also to the rate of change in neutron density to automatically maintain the total power output at a substantially constant predetermined value. A safety rod means responsive to neutron density is also provided for keeping the power output below a predetermined maximum value at all times.

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

NASA Astrophysics Data System (ADS)

Schneider, E. A.; Deinert, M. R.; Cady, K. B.

2006-10-01

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Updated methodology for nuclear magnetic resonance characterization of shales

NASA Astrophysics Data System (ADS)

Washburn, Kathryn E.; Birdwell, Justin E.

2013-08-01

Unconventional petroleum resources, particularly in shales, are expected to play an increasingly important role in the world's energy portfolio in the coming years. Nuclear magnetic resonance (NMR), particularly at low-field, provides important information in the evaluation of shale resources. Most of the low-field NMR analyses performed on shale samples rely heavily on standard T1 and T2 measurements. We present a new approach using solid echoes in the measurement of T1 and T1-T2 correlations that addresses some of the challenges encountered when making NMR measurements on shale samples compared to conventional reservoir rocks. Combining these techniques with standard T1 and T2 measurements provides a more complete assessment of the hydrogen-bearing constituents (e.g., bitumen, kerogen, clay-bound water) in shale samples. These methods are applied to immature and pyrolyzed oil shale samples to examine the solid and highly viscous organic phases present during the petroleum generation process. The solid echo measurements produce additional signal in the oil shale samples compared to the standard methodologies, indicating the presence of components undergoing homonuclear dipolar coupling. The results presented here include the first low-field NMR measurements performed on kerogen as well as detailed NMR analysis of highly viscous thermally generated bitumen present in pyrolyzed oil shale.

Updated methodology for nuclear magnetic resonance characterization of shales

USGS Publications Warehouse

Washburn, Kathryn E.; Birdwell, Justin E.

2013-01-01

Unconventional petroleum resources, particularly in shales, are expected to play an increasingly important role in the world’s energy portfolio in the coming years. Nuclear magnetic resonance (NMR), particularly at low-field, provides important information in the evaluation of shale resources. Most of the low-field NMR analyses performed on shale samples rely heavily on standard T1 and T2 measurements. We present a new approach using solid echoes in the measurement of T1 and T1–T2 correlations that addresses some of the challenges encountered when making NMR measurements on shale samples compared to conventional reservoir rocks. Combining these techniques with standard T1 and T2 measurements provides a more complete assessment of the hydrogen-bearing constituents (e.g., bitumen, kerogen, clay-bound water) in shale samples. These methods are applied to immature and pyrolyzed oil shale samples to examine the solid and highly viscous organic phases present during the petroleum generation process. The solid echo measurements produce additional signal in the oil shale samples compared to the standard methodologies, indicating the presence of components undergoing homonuclear dipolar coupling. The results presented here include the first low-field NMR measurements performed on kerogen as well as detailed NMR analysis of highly viscous thermally generated bitumen present in pyrolyzed oil shale.

Effect of milling conditions on solid-state amorphization of glipizide, and characterization and stability of solid forms.

PubMed

Xu, Kailin; Xiong, Xinnuo; Zhai, Yuanming; Wang, Lili; Li, Shanshan; Yan, Jin; Wu, Di; Ma, Xiaoli; Li, Hui

2016-09-10

In this study, the amorphization of glipizide was systematically investigated through high-energy ball milling at different temperatures. The results of solid-state amorphization through milling indicated that glipizide underwent direct crystal-to-glass transformation at 15 and 25°C and crystal-to-glass-to-crystal conversion at 35°C; hence, milling time and temperature had significant effects on the amorphization of glipizide, which should be effectively controlled to obtain totally amorphous glipizide. Solid forms of glipizide were detailedly characterized through analyses of X-ray powder diffraction, morphology, thermal curves, vibrational spectra, and solid-state nuclear magnetic resonance. The physical stability of solid forms was investigated under different levels of relative humidity (RH) at 25°C. Forms I and III are kinetically stable and do not form any new solid-state forms at various RH levels. By contrast, Form II is kinetically unstable, undergoing direct glass-to-crystal transformation when RH levels higher than 32.8%. Therefore, stability investigation indicated that Form II should be stored under relatively dry conditions to prevent rapid crystallization. High temperatures can also induce the solid-state transformation of Form II; the conversion rate increased with increasing temperature. Copyright © 2016 Elsevier B.V. All rights reserved.

Spin Noise Detection of Nuclear Hyperpolarization at 1.2 K

PubMed Central

Pöschko, Maria Theresia; Vuichoud, Basile; Milani, Jonas; Bornet, Aurélien; Bechmann, Matthias; Bodenhausen, Geoffrey; Jannin, Sami; Müller, Norbert

2015-01-01

We report proton spin noise spectra of a hyperpolarized solid sample of commonly used “DNP (dynamic nuclear polarization) juice” containing TEMPOL (4-hydroxy-2,2,6,6-tetramethylpiperidine N-oxide) and irradiated by a microwave field at a temperature of 1.2 K in a magnetic field of 6.7 T. The line shapes of the spin noise power spectra are sensitive to the variation of the microwave irradiation frequency and change from dip to bump, when the electron Larmor frequency is crossed, which is shown to be in good accordance with theory by simulations. Small but significant deviations from these predictions are observed, which can be related to spin noise and radiation damping phenomena that have been reported in thermally polarized systems. The non-linear dependence of the spin noise integral on nuclear polarization provides a means to monitor hyperpolarization semi-quantitatively without any perturbation of the spin system by radio frequency irradiation. PMID:26477605

On fast solid-body rotation of the solar core and differential (liquid-like) rotation of the solar surface

NASA Astrophysics Data System (ADS)

Pashitskii, E. A.

2017-07-01

On the basis of a two-component (two-fluid) hydrodynamic model, it is shown that the probable phenomenon of solar core rotation with a velocity higher than the average velocity of global rotation of the Sun, discovered by the SOHO mission, can be related to fast solid-body rotation of the light hydrogen component of the solar plasma, which is caused by thermonuclear fusion of hydrogen into helium inside the hot dense solar core. Thermonuclear fusion of four protons into a helium nucleus (α-particle) creates a large free specific volume per unit particle due to the large difference between the densities of the solar plasma and nuclear matter. As a result, an efficient volumetric sink of one of the components of the solar substance—hydrogen—forms inside the solar core. Therefore, a steady-state radial proton flux converging to the center should exist inside the Sun, which maintains a constant concentration of hydrogen as it burns out in the solar core. It is demonstrated that such a converging flux of hydrogen plasma with the radial velocity v r ( r) = -β r creates a convective, v r ∂ v φ/∂ r, and a local Coriolis, v r v φ/ r,φ nonlinear hydrodynamic forces in the solar plasma, rotating with the azimuthal velocity v φ. In the absence of dissipation, these forces should cause an exponential growth of the solid-body rotation velocity of the hydrogen component inside the solar core. However, friction between the hydrogen and helium components of the solar plasma due to Coulomb collisions of protons with α-particles results in a steady-state regime of rotation of the hydrogen component in the solar core with an angular velocity substantially exceeding the global rotational velocity of the Sun. It is suggested that the observed differential (liquid-like) rotation of the visible surface of the Sun (photosphere) with the maximum angular velocity at the equator is caused by sold-body rotation of the solar plasma in the radiation zone and strong turbulence in the tachocline layer, where the turbulent viscosity reaches its maximum value at the equator. There, the tachocline layer exerts the most efficient drag on the less dense outer layers of the solar plasma, which are slowed down due to the interaction with the ambient space plasma (solar wind).

Lunar Fluid Core and Solid-Body Tides

NASA Technical Reports Server (NTRS)

Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

2005-01-01

Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2-5] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening has been improving [3,5] and now seems significant. This strengthens the case for a fluid lunar core.

The Effect of Core Configuration on Thermal Barrier Thermal Performance

NASA Technical Reports Server (NTRS)

DeMange, Jeffrey J.; Bott, Robert H.; Druesedow, Anne S.

2015-01-01

Thermal barriers and seals are integral components in the thermal protection systems (TPS) of nearly all aerospace vehicles. They are used to minimize heat transfer through interfaces and gaps and protect underlying temperature-sensitive components. The core insulation has a significant impact on both the thermal and mechanical properties of compliant thermal barriers. Proper selection of an appropriate core configuration to mitigate conductive, convective and radiative heat transfer through the thermal barrier is challenging. Additionally, optimization of the thermal barrier for thermal performance may have counteracting effects on mechanical performance. Experimental evaluations have been conducted to better understand the effect of insulation density on permeability and leakage performance, which can significantly impact the resistance to convective heat transfer. The effect of core density on mechanical performance was also previously investigated and will be reviewed. Simple thermal models were also developed to determine the impact of various core parameters on downstream temperatures. An extended understanding of these factors can improve the ability to design and implement these critical TPS components.

Mechanochemical Synthesis of Li2MnO3 Shell/LiMO2 (M = Ni, Co, Mn) Core-Structured Nanocomposites for Lithium-Ion Batteries

PubMed Central

Noh, Jae-Kyo; Kim, Soo; Kim, Haesik; Choi, Wonchang; Chang, Wonyoung; Byun, Dongjin; Cho, Byung-Won; Chung, Kyung Yoon

2014-01-01

Core/shell-like nanostructured xLi2MnO3·(1−x)LiMO2 (M = Ni, Co, Mn) composite cathode materials are successfully synthesized through a simple solid-state reaction using a mechanochemical ball-milling process. The LiMO2 core is designed to have a high-content of Ni, which increases the specific capacity. The detrimental surface effects arising from the high Ni-content are countered by the Li2MnO3 shell, which stabilizes the nanoparticles. The electrochemical performances and thermal stabilities of the synthesized nanocomposites are compared with those of bare LiMO2. In particular, the results of time-resolved X-ray diffraction (TR-XRD) analyses of xLi2MnO3·(1−x)LiMO2 nanocomposites as well as their differential scanning calorimetry (DSC) profiles demonstrate that the Li2MnO3 shell is effective in stabilizing the LiMO2 core at high temperatures, making the nanocomposites highly suitable from a safety viewpoint. PMID:24784478

Dry-air-stable lithium silicide-lithium oxide core-shell nanoparticles as high-capacity prelithiation reagents.

PubMed

Zhao, Jie; Lu, Zhenda; Liu, Nian; Lee, Hyun-Wook; McDowell, Matthew T; Cui, Yi

2014-10-03

Rapid progress has been made in realizing battery electrode materials with high capacity and long-term cyclability in the past decade. However, low first-cycle Coulombic efficiency as a result of the formation of a solid electrolyte interphase and Li trapping at the anodes, remains unresolved. Here we report LixSi-Li2O core-shell nanoparticles as an excellent prelithiation reagent with high specific capacity to compensate the first-cycle capacity loss. These nanoparticles are produced via a one-step thermal alloying process. LixSi-Li2O core-shell nanoparticles are processible in a slurry and exhibit high capacity under dry-air conditions with the protection of a Li2O passivation shell, indicating that these nanoparticles are potentially compatible with industrial battery fabrication processes. Both Si and graphite anodes are successfully prelithiated with these nanoparticles to achieve high first-cycle Coulombic efficiencies of 94% to >100%. The LixSi-Li2O core-shell nanoparticles enable the practical implementation of high-performance electrode materials in lithium-ion batteries.

Top-down freezing in a Fe-FeS core and Ganymede's present-day magnetic field

NASA Astrophysics Data System (ADS)

Rückriemen, Tina; Breuer, Doris; Spohn, Tilman

2018-06-01

Ganymede's core most likely possesses an active dynamo today, which produces a magnetic field at the surface of ∼ 719 nT. Thermochemical convection triggered by cooling of the core is a feasible power source for the dynamo. Experiments of different research groups indicate low pressure gradients of the melting temperatures for Fe-FeS core alloys at pressures prevailing in Ganymede's core ( < 10 GPa). This may entail that the core crystallizes from the top instead of from the bottom as is expected for Earth's core. Depending on the core sulfur concentration being more iron- or more sulfur-rich than the eutectic concentration either snowing iron crystals or a solid FeS layer can form at the top of the core. We investigate whether these two core crystallization scenarios are capable of explaining Ganymede's present magnetic activity. To do so, we set up a parametrized one-dimensional thermal evolution model. We explore a wide range of parameters by running a large set of Monte Carlo simulations. Both freezing scenarios can explain Ganymede's present-day magnetic field. Dynamos of iron snow models are rather young ( < 1 Gyr), whereas dynamos below the FeS layer can be both young and much older ( ∼ 3.8 Gyr). Successful models preferably contain less radiogenic heat sources in the mantle than the chondritic abundance and show a correlation between the reference viscosity in the mantle and the initial core sulfur concentration.

Multi-Group Formulation of the Temperature-Dependent Resonance Scattering Model and its Impact on Reactor Core Parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghrayeb, Shadi Z.; Ougouag, Abderrafi M.; Ouisloumen, Mohamed

2014-01-01

A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects, stemming from lattice atoms thermal motion and accounts for it within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler Reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering,more » which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to -10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes the results done on this topic to date.« less

The Thermal Conductivity of Earth's Core: A Key Geophysical Parameter's Constraints and Uncertainties

NASA Astrophysics Data System (ADS)

Williams, Q.

2018-05-01

The thermal conductivity of iron alloys at high pressures and temperatures is a critical parameter in governing ( a) the present-day heat flow out of Earth's core, ( b) the inferred age of Earth's inner core, and ( c) the thermal evolution of Earth's core and lowermost mantle. It is, however, one of the least well-constrained important geophysical parameters, with current estimates for end-member iron under core-mantle boundary conditions varying by about a factor of 6. Here, the current state of calculations, measurements, and inferences that constrain thermal conductivity at core conditions are reviewed. The applicability of the Wiedemann-Franz law, commonly used to convert electrical resistivity data to thermal conductivity data, is probed: Here, whether the constant of proportionality, the Lorenz number, is constant at extreme conditions is of vital importance. Electron-electron inelastic scattering and increases in Fermi-liquid-like behavior may cause uncertainties in thermal conductivities derived from both first-principles-associated calculations and electrical conductivity measurements. Additional uncertainties include the role of alloying constituents and local magnetic moments of iron in modulating the thermal conductivity. Thus, uncertainties in thermal conductivity remain pervasive, and hence a broad range of core heat flows and inner core ages appear to remain plausible.

Lunar Science from Lunar Laser Ranging

NASA Technical Reports Server (NTRS)

Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

2013-01-01

Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, tidal Love number k2, and moment of inertia differences. There is weaker sensitivity to flattening of the core/mantle boundary (CMB) and fluid core moment of inertia. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to variations in lunar rotation, orientation and tidal displacements. Past solutions using the LLR data have given results for Love numbers plus dissipation due to solid-body tides and fluid core. Detection of the fluid core polar minus equatorial moment of inertia difference due to CMB flattening is weakly significant. This strengthens the case for a fluid lunar core. Future approaches are considered to detect a solid inner core.

Implications of Convection in the Moon and the Terrestrial Planets

NASA Technical Reports Server (NTRS)

Turcotte, D. L.

1985-01-01

The early evolution of the Moon and its implications for the early evolution of the Earth was studied. The study is divided into two parts: (1) studies of core formation. Cosmochemical studies strongly favor a near-homogeneous accretion of the Earth. It is shown that core segregation probably occurred within the first 10,000 years of Earth history. It is found that dissipative heating may be a viable mechanism for core segregation if sufficiently large bodies of liquid iron can form; (2) early thermal evolution of the Earth and Moon. The energy associated with the accretion of the Earth and the segregation of the core is more than sufficient to melt the entire Earth. The increase in the mantle liquidus with depth (pressure) is the dominant effect influencing heat transfer through the magma ocean. It is found that a magma ocean with a depth of 100 km would have existed as the Earth accreted. It is concluded that this magma ocean zone refined the earth resulting in the simultaneous formation of the core and the atmosphere during accretion. The resulting mantle was a well-mixed solid with a near pyrolite composition.

Identification of a Functional, CRM-1-Dependent Nuclear Export Signal in Hepatitis C Virus Core Protein

PubMed Central

Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

2011-01-01

Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified. We show here that the aa(109–133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1–173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication. Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection. PMID:22039426

Planetary cores, their energy flux relationship, and its implications

NASA Astrophysics Data System (ADS)

Johnson, Fred M.

2018-02-01

Integrated surface heat flux data from each planet in our solar system plus over 50 stars, including our Sun, was plotted against each object's known mass to generate a continuous exponential curve at an R-squared value of 0.99. The unexpected yet undeniable implication of this study is that all planets and celestial objects have a similar mode of energy production. It is widely accepted that proton-proton reactions require hydrogen gas at temperatures of about 15 million degrees, neither of which can plausibly exist inside a terrestrial planet. Hence, this paper proposes a nuclear fission mechanism for all luminous celestial objects, and uses this mechanism to further suggest a developmental narrative for all celestial bodies, including our Sun. This narrative was deduced from an exponential curve drawn adjacent to the first and passing through the Earth's solid core (as a known prototype). This trend line was used to predict the core masses for each planet as a function of its luminosity.

In-vivo imaging of nanoshell extravasation from solid tumor vasculature by photoacoustic microscopy

NASA Astrophysics Data System (ADS)

Li, Meng-Lin; Schwartz, Jon A.; Wang, James; Stoica, George; Wang, Lihong V.

2007-02-01

In this study, high resolution reflection-mode (backward-mode) photoacoustic microscopy (PAM) is used to noninvasively image progressive extravasation and accumulation of nanoshells within a solid tumor in vivo. This study takes advantage of the strong near-infrared absorption of nanoshells, a novel type of optically tunable gold nanoparticles that tend to extravasate from leaky tumor vasculatures (i.e., passive targeting) via the "enhanced permeability and retention" effect due to their nanoscale size. Tumors were grown in immunocompetent BALB/c mice by subcutaneous inoculation of CT26.wt murine colon carcinoma cells. PEGylated nanoshells with a peak optical absorption at ~800 nm were intravenously administered. Pre-scans prior to nanoshell injection were taken using a 584-nm laser source to highlight blood content and an 800-nm laser source to mark the background limit for nanoshell accumulation. After injection, the three-dimensional nanoshell distribution inside the tumor was monitored by PAM for 7 hours. Experimental results show that nanoshell accumulation is heterogeneous in tumors: more concentrated within the tumor cortex and largely absent from the tumor core. This correlates with others' observation that drug delivery within tumor cores is ineffective because of both high interstitial pressure and tendency to necrosis of tumor cores. Since nanoshells have been recently applied to thermal therapy for subcutaneous tumors, we anticipate that PAM will be important to this therapeutic technique.

Supercontinuum generation covering the entire 0.4-5 µm transmission window in a tapered ultra-high numerical aperture all-solid fluorotellurite fiber

NASA Astrophysics Data System (ADS)

Jia, Z. X.; Yao, C. F.; Jia, S. J.; Wang, F.; Wang, S. B.; Zhao, Z. P.; Liao, M. S.; Qin, G. S.; Hu, L. L.; Ohishi, Y.; Qin, W. P.

2018-02-01

Enormous efforts have been made to realize supercontinuum (SC) generation covering the entire transmission window of fiber materials for their wide applications in many fields. Here we demonstrate ultra-broadband SC generation from 400 to 5140 nm in a tapered ultra-high numerical aperture (NA) all-solid fluorotellurite fiber pumped by a 1560 nm mode-locked fiber laser. The fluorotellurite fibers are fabricated using a rod-in-tube method. The core and cladding materials are TeO2-BaF2-Y2O3- and TeO2-modified fluoroaluminate glasses, respectively, which have large refractive index contrast and similar thermal expansion coefficients and softening temperatures. The NA at 3200 nm of the fluorotellurite fiber is about 1.11. Furthermore, tapered fluorotellurite fibers are prepared using an elongation machine. SC generation covering the entire 0.4-5 µm transmission window is achieved in a tapered fluorotellurite fiber for a pumping peak power of ~10.5 kW through synergetic control of dispersion, nonlinearity, confinement loss and other unexpected effects (e.g. the attachment of dust or water to the surface of the fiber core) of the fiber. Our results show that tapered ultra-high NA all-solid soft glass fibers have a potential for generating SC light covering their entire transmission window.

Temperature effect on laser-induced breakdown spectroscopy spectra of molten and solid salts

NASA Astrophysics Data System (ADS)

Hanson, Cynthia; Phongikaroon, Supathorn; Scott, Jill R.

2014-07-01

Laser-induced breakdown spectroscopy (LIBS) has been investigated as a potential analytical tool to improve operations and safeguards for electrorefiners, such as those used in processing spent nuclear fuel. This study set out to better understand the effect of sample temperature and physical state on LIBS spectra of molten and solid salts by building calibration curves of cerium and assessing self-absorption, plasma temperature, electron density, and local thermal equilibrium (LTE). Samples were composed of a LiCl-KCl eutectic salt, an internal standard of MnCl2, and varying concentrations of CeCl3 (0.1, 0.3, 0.5, 0.8, and 1.0 wt.% Ce) under different temperatures (773, 723, 673, 623, and 573 K). Analysis of salts in their molten form is preferred as plasma plumes from molten samples experienced less self-absorption, less variability in plasma temperature, and higher clearance of the minimum electron density required for local thermal equilibrium. These differences are attributed to plasma dynamics as a result of phase changes. Spectral reproducibility was also better in the molten state due to sample homogeneity.

Preparation and Physical Characterization of a Diclofenac-Ranitidine Co-precipitate for Improving the Dissolution of Diclofenac.

PubMed

Gaitano, Robertino O; Calvo, Natalia L; Narda, Griselda E; Kaufman, Teodoro S; Maggio, Rubén M; Brusau, Elena V

2016-03-01

Mixing aqueous solutions of sodium diclofenac (DIC-Na) and ranitidine hydrochloride (RAN·HCl) afforded an off-white solid (DIC-RAN) that was investigated from the microscopic, thermal, diffractometric, spectroscopic, and functional (chemometrics-assisted dissolution) points of view. The solid has a 2:1 (DIC:RAN) molar ratio according to (1)H nuclear magnetic resonance spectroscopy. It is thermally stable, displaying a broad endothermic signal centered at 105°C in the thermogram, and its characteristic reflections in the powder X-ray diffractogram remained unchanged after a 3-month aging period. Scanning electron microscopy micrographs uncovered its morphology, whereas the spectral data suggested an interaction between the carboxylic acid of DIC and the alkyldimethylamino moiety of RAN. The dissolution of DIC-RAN was monitored at different pH values by an ultraviolet/chemometrics procedure, being complete within 5 min at pH 6.8. This compares favorably with the dissolution of a DIC-Na sample of the same particle size. Copyright © 2016 American Pharmacists Association®. Published by Elsevier Inc. All rights reserved.

Affordable Development and Demonstration of a Small NTR Engine and Stage: How Small is Big Enough?

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Sefcik, Robert J.; Fittje, James E.; McCurdy, David R.; Qualls, Arthur L.; Schnitzler, Bruce G.; Werner, James E.; Weitzberg (Abraham); Joyner, Claude R.

2015-01-01

The Nuclear Thermal Rocket (NTR) derives its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core. It generates high thrust and has a specific impulse potential of approximately 900 seconds - a 100% increase over today's best chemical rockets. The Nuclear Thermal Propulsion (NTP) project, funded by NASA's AES program, includes five key task activities: (1) Recapture, demonstration, and validation of heritage graphite composite (GC) fuel (selected as the "Lead Fuel" option); (2) Engine Conceptual Design; (3) Operating Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable Development Strategy. During FY'14, a preliminary DDT&E plan and schedule for NTP development was outlined by GRC, DOE and industry that involved significant system-level demonstration projects that included GTD tests at the NNSS, followed by a FTD mission. To reduce cost for the GTD tests and FTD mission, small NTR engines, in either the 7.5 or 16.5 klbf thrust class, were considered. Both engine options used GC fuel and a "common" fuel element (FE) design. The small approximately 7.5 klbf "criticality-limited" engine produces approximately 157 megawatts of thermal power (MWt) and its core is configured with parallel rows of hexagonal-shaped FEs and tie tubes (TTs) with a FE to TT ratio of approximately 1:1. The larger approximately 16.5 klbf Small Nuclear Rocket Engine (SNRE), developed by LANL at the end of the Rover program, produces approximately 367 MWt and has a FE to TT ratio of approximately 2:1. Although both engines use a common 35 inch (approximately 89 cm) long FE, the SNRE's larger diameter core contains approximately 300 more FEs needed to produce an additional 210 MWt of power. To reduce the cost of the FTD mission, a simple "1-burn" lunar flyby mission was considered to reduce the LH2 propellant loading, the stage size and complexity. Use of existing and flight proven liquid rocket and stage hardware (e.g., from the RL10B-2 engine and Delta Cryogenic Second Stage) was also maximized to further aid affordability. This paper examines the pros and cons of using these two small engine options, including their potential to support future human exploration missions to the Moon, near Earth asteroids, and Mars, and recommends a preferred size. It also provides a preliminary assessment of the key activities, development options, and schedule required to affordably build, ground test and fly a small NTR engine and stage within a 10-year timeframe.

Impact of Americium-241 (n,γ) Branching Ratio on SFR Core Reactivity and Spent Fuel Characteristics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hiruta, Hikaru; Youinou, Gilles J.; Dixon, Brent W.

An accurate prediction of core physics and fuel cycle parameters largely depends on the order of details and accuracy in nuclear data taken into account for actual calculations. 241Am is a major gateway nuclide for most of minor actinides and thus important nuclide for core physics and fuel-cycle calculations. The 241Am(n,?) branching ratio (BR) is in fact the energy dependent (see Fig. 1), therefore, it is necessary to taken into account the spectrum effect on the calculation of the average BR for the full-core depletion calculations. Moreover, the accuracy of the BR used in the depletion calculations could significantly influencemore » the core physics performance and post irradiated fuel compositions. The BR of 241Am(n,?) in ENDF/B-VII.0 library is relatively small and flat in thermal energy range, gradually increases within the intermediate energy range, and even becomes larger at the fast energy range. This indicates that the properly collapsed BR for fast reactors could be significantly different from that of thermal reactors. The evaluated BRs are also differ from one evaluation to another. As seen in Table I, average BRs for several evaluated libraries calculated by means of a fast spectrum are similar but have some differences. Most of currently available depletion codes use a pre-determined single value BR for each library. However, ideally it should be determined on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different 241Am(n,?) BRs on core and spent fuel parameters. This paper investigates the impact of the 241Am(n,?) BR on the results of SFR full-core based fuel-cycle calculations. The analysis is performed by gradually increasing the value of BR from 0.15 to 0.25 and studying its impact on the core reactivity and characteristics of SFR spent fuels over extended storage times (~10,000 years).« less

Organic petrology of the Aptian-age section in the downdip Mississippi Interior Salt Basin, Mississippi, USA: Observations and preliminary implications for thermal maturation history

USGS Publications Warehouse

Valentine, Brett J.; Hackley, Paul C.; Enomoto, Catherine B.; Bove, Alana M.; Dulong, Frank T.; Lohr, Celeste D.; Scott, Krystina R.

2014-01-01

This study identifies a thermal maturity anomaly within the downdip Mississippi Interior Salt Basin (MISB) of southern Mississippi, USA, through examination of bitumen reflectance data from Aptian-age strata (Sligo Formation, Pine Island Shale, James Limestone, and Rodessa Formation). U.S. Geological Survey (USGS) reconnaissance investigations conducted in 2011–2012 examined Aptian-age thermal maturity trends across the onshore northern Gulf of Mexico region and indicated that the section in the downdip MISB is approaching the wet gas/condensate window (Ro~1.2%). A focused study in 2012–2013 used 6 whole core, one sidewall core, and 49 high-graded cutting samples (depth range of 13,000–16,500 ft [3962.4–5029.2 m] below surface) collected from 15 downdip MISB wells for mineralogy, fluid inclusion, organic geochemistry, and organic petrographic analysis. Based on native solid bitumen reflectance (Ro generally > 0.8%; interpreted to be post-oil indigenous bitumens matured in situ), Ro values increase regionally across the MISB from the southeast to the northwest. Thermal maturity in the eastern half of the basin (Ro range 1.0 to 1.25%) appears to be related to present-day burial depth and shows a gradual increase with respect to depth. To the west, thermal maturity continues to increase even as the Aptian section shallows structurally on the Adams County High (Ro range 1.4 to > 1.8%). After evaluating the possible thermal agents responsible for increasing maturity at shallower depths (i.e., igneous activity, proximity to salt, variations in regional heat flux, and uplift), we tentatively propose that either greater paleoheat flow or deeper burial coupled with uplift in the western part of the MISB could be responsible for the thermal maturity anomaly. Further research and additional data are needed to determine the cause(s) of the thermal anomaly.

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

THE LAST MINUTES OF OXYGEN SHELL BURNING IN A MASSIVE STAR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Müller, Bernhard; Viallet, Maxime; Janka, Hans-Thomas

We present the first  4 π– three-dimensional (3D) simulation of the last minutes of oxygen shell burning in an 18 M {sub ⊙} supernova progenitor up to the onset of core collapse. A moving inner boundary is used to accurately model the contraction of the silicon and iron core according to a one-dimensional stellar evolution model with a self-consistent treatment of core deleptonization and nuclear quasi-equilibrium. The simulation covers the full solid angle to allow the emergence of large-scale convective modes. Due to core contraction and the concomitant acceleration of nuclear burning, the convective Mach number increases to ∼0.1 at collapse,more » and an ℓ  = 2 mode emerges shortly before the end of the simulation. Aside from a growth of the oxygen shell from 0.51 M {sub ⊙} to 0.56 M {sub ⊙} due to entrainment from the carbon shell, the convective flow is reasonably well described by mixing-length theory, and the dominant scales are compatible with estimates from linear stability analysis. We deduce that artificial changes in the physics, such as accelerated core contraction, can have precarious consequences for the state of convection at collapse. We argue that scaling laws for the convective velocities and eddy sizes furnish good estimates for the state of shell convection at collapse and develop a simple analytic theory for the impact of convective seed perturbations on shock revival in the ensuing supernova. We predict a reduction of the critical luminosity for explosion by 12% – 24% due to seed asphericities for our 3D progenitor model relative to the case without large seed perturbations.« less

A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D’Auria, F; Rohatgi, Upendra S.

The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

Habitable evaporated cores: transforming mini-Neptunes into super-Earths in the habitable zones of M dwarfs.

PubMed

Luger, R; Barnes, R; Lopez, E; Fortney, J; Jackson, B; Meadows, V

2015-01-01

We show that photoevaporation of small gaseous exoplanets ("mini-Neptunes") in the habitable zones of M dwarfs can remove several Earth masses of hydrogen and helium from these planets and transform them into potentially habitable worlds. We couple X-ray/extreme ultraviolet (XUV)-driven escape, thermal evolution, tidal evolution, and orbital migration to explore the types of systems that may harbor such "habitable evaporated cores" (HECs). We find that HECs are most likely to form from planets with ∼1 M⊕ solid cores with up to about 50% H/He by mass, though whether or not a given mini-Neptune forms a HEC is highly dependent on the early XUV evolution of the host star. As terrestrial planet formation around M dwarfs by accumulation of local material is likely to form planets that are small and dry, evaporation of small migrating mini-Neptunes could be one of the dominant formation mechanisms for volatile-rich Earths around these stars.

Aluminum integral foams with tailored density profile by adapted blowing agents

NASA Astrophysics Data System (ADS)

Hartmann, Johannes; Fiegl, Tobias; Körner, Carolin

2014-05-01

The goal of the present work is the variation of the structure of aluminum integral foams regarding the thickness of the integral solid skin as well as the density profile. A modified die casting process, namely integral foam molding, is used in which an aluminum melt and blowing agent particles (magnesium hydride MgH2) are injected in a permanent steel mold. The high solidification rates at the cooled walls of the mold lead to the formation of a solid skin. In the inner region, hydrogen is released by thermal decomposition of MgH2 particles. Thus, the pore formation takes place parallel to the continuing solidification of the melt. The thickness of the solid skin and the density profile of the core strongly depend on the interplay between solidification velocity and kinetics of hydrogen release. By varying the melt and blowing agent properties, the structure of integral foams can be systematically changed to meet the requirements of the desired field of application of the produced component.

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

Online monitoring of the Osiris reactor with the Nucifer neutrino detector

NASA Astrophysics Data System (ADS)

Boireau, G.; Bouvet, L.; Collin, A. P.; Coulloux, G.; Cribier, M.; Deschamp, H.; Durand, V.; Fechner, M.; Fischer, V.; Gaffiot, J.; Gérard Castaing, N.; Granelli, R.; Kato, Y.; Lasserre, T.; Latron, L.; Legou, P.; Letourneau, A.; Lhuillier, D.; Mention, G.; Mueller, Th. A.; Nghiem, T.-A.; Pedrol, N.; Pelzer, J.; Pequignot, M.; Piret, Y.; Prono, G.; Scola, L.; Starzinski, P.; Vivier, M.; Dumonteil, E.; Mancusi, D.; Varignon, C.; Buck, C.; Lindner, M.; Bazoma, J.; Bouvier, S.; Bui, V. M.; Communeau, V.; Cucoanes, A.; Fallot, M.; Gautier, M.; Giot, L.; Guilloux, G.; Lenoir, M.; Martino, J.; Mercier, G.; Milleto, T.; Peuvrel, N.; Porta, A.; Le Quéré, N.; Renard, C.; Rigalleau, L. M.; Roy, D.; Vilajosana, T.; Yermia, F.; Nucifer Collaboration

2016-06-01

Originally designed as a new nuclear reactor monitoring device, the Nucifer detector has successfully detected its first neutrinos. We provide the second-shortest baseline measurement of the reactor neutrino flux. The detection of electron antineutrinos emitted in the decay chains of the fission products, combined with reactor core simulations, provides a new tool to assess both the thermal power and the fissile content of the whole nuclear core and could be used by the International Agency for Atomic Energy to enhance the safeguards of civil nuclear reactors. Deployed at only 7.2 m away from the compact Osiris research reactor core (70 MW) operating at the Saclay research center of the French Alternative Energies and Atomic Energy Commission, the experiment also exhibits a well-suited configuration to search for a new short baseline oscillation. We report the first results of the Nucifer experiment, describing the performances of the ˜0.85 m3 detector remotely operating at a shallow depth equivalent to ˜12 m of water and under intense background radiation conditions. Based on 145 (106) days of data with the reactor on (off), leading to the detection of an estimated 40760 ν¯ e , the mean number of detected antineutrinos is 281 ±7 (stat )±18 (syst )ν¯ e/day , in agreement with the prediction of 277 ±23 ν¯ e/day . Because of the large background, no conclusive results on the existence of light sterile neutrinos could be derived, however. As a first societal application we quantify how antineutrinos could be used for the Plutonium Management and Disposition Agreement.

SiC/SiC Cladding Materials Properties Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snead, Mary A.; Katoh, Yutai; Koyanagi, Takaaki

When a new class of material is considered for a nuclear core structure, the in-pile performance is usually assessed based on multi-physics modeling in coordination with experiments. This report aims to provide data for the mechanical and physical properties and environmental resistance of silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites for use in modeling for their application as accidenttolerant fuel cladding for light water reactors (LWRs). The properties are specific for tube geometry, although many properties can be predicted from planar specimen data. This report presents various properties, including mechanical properties, thermal properties, chemical stability under normal and offnormalmore » operation conditions, hermeticity, and irradiation resistance. Table S.1 summarizes those properties mainly for nuclear-grade SiC/SiC composites fabricated via chemical vapor infiltration (CVI). While most of the important properties are available, this work found that data for the in-pile hydrothermal corrosion resistance of SiC materials and for thermal properties of tube materials are lacking for evaluation of SiC-based cladding for LWR applications.« less

Radiation tolerance of piezoelectric bulk single-crystal aluminum nitride

DOE Office of Scientific and Technical Information (OSTI.GOV)

David A. Parks; Bernhard R. Tittmann

2014-07-01

For practical use in harsh radiation environments, we pose selection criteria for piezoelectric materials for nondestructive evaluation (NDE) and material characterization. Using these criteria, piezoelectric aluminum nitride is shown to be an excellent candidate. The results of tests on an aluminumnitride-based transducer operating in a nuclear reactor are also presented. We demonstrate the tolerance of single-crystal piezoelectric aluminum nitride after fast and thermal neutron fluences of 1.85 × 1018 neutron/cm2 and 5.8 × 1018 neutron/cm2, respectively, and a gamma dose of 26.8 MGy. The radiation hardness of AlN is most evident from the unaltered piezoelectric coefficient d33, which measured 5.5more » pC/N after a fast and thermal neutron exposure in a nuclear reactor core for over 120 MWh, in agreement with the published literature value. The results offer potential for improving reactor safety and furthering the understanding of radiation effects on materials by enabling structural health monitoring and NDE in spite of the high levels of radiation and high temperatures, which are known to destroy typical commercial ultrasonic transducers.« less

The COPERNIC3 project: how AREVA is successfully developing an advanced global fuel rod performance code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garnier, Ch.; Mailhe, P.; Sontheimer, F.

2007-07-01

Fuel performance is a key factor for minimizing operating costs in nuclear plants. One of the important aspects of fuel performance is fuel rod design, based upon reliable tools able to verify the safety of current fuel solutions, prevent potential issues in new core managements and guide the invention of tomorrow's fuels. AREVA is developing its future global fuel rod code COPERNIC3, which is able to calculate the thermal-mechanical behavior of advanced fuel rods in nuclear plants. Some of the best practices to achieve this goal are described, by reviewing the three pillars of a fuel rod code: the database,more » the modelling and the computer and numerical aspects. At first, the COPERNIC3 database content is described, accompanied by the tools developed to effectively exploit the data. Then is given an overview of the main modelling aspects, by emphasizing the thermal, fission gas release and mechanical sub-models. In the last part, numerical solutions are detailed in order to increase the computational performance of the code, with a presentation of software configuration management solutions. (authors)« less

Control of the Speed of a Light-Induced Spin Transition through Mesoscale Core-Shell Architecture.

PubMed

Felts, Ashley C; Slimani, Ahmed; Cain, John M; Andrus, Matthew J; Ahir, Akhil R; Abboud, Khalil A; Meisel, Mark W; Boukheddaden, Kamel; Talham, Daniel R

2018-05-02

The rate of the light-induced spin transition in a coordination polymer network solid dramatically increases when included as the core in mesoscale core-shell particles. A series of photomagnetic coordination polymer core-shell heterostructures, based on the light-switchable Rb a Co b [Fe(CN) 6 ] c · mH 2 O (RbCoFe-PBA) as core with the isostructural K j Ni k [Cr(CN) 6 ] l · nH 2 O (KNiCr-PBA) as shell, are studied using temperature-dependent powder X-ray diffraction and SQUID magnetometry. The core RbCoFe-PBA exhibits a charge transfer-induced spin transition (CTIST), which can be thermally and optically induced. When coupled to the shell, the rate of the optically induced transition from low spin to high spin increases. Isothermal relaxation from the optically induced high spin state of the core back to the low spin state and activation energies associated with the transition between these states were measured. The presence of a shell decreases the activation energy, which is associated with the elastic properties of the core. Numerical simulations using an electro-elastic model for the spin transition in core-shell particles supports the findings, demonstrating how coupling of the core to the shell changes the elastic properties of the system. The ability to tune the rate of optically induced magnetic and structural phase transitions through control of mesoscale architecture presents a new approach to the development of photoswitchable materials with tailored properties.

High-resolution magnetic resonance spectroscopy using a solid-state spin sensor

NASA Astrophysics Data System (ADS)

Glenn, David R.; Bucher, Dominik B.; Lee, Junghyun; Lukin, Mikhail D.; Park, Hongkun; Walsworth, Ronald L.

2018-03-01

Quantum systems that consist of solid-state electronic spins can be sensitive detectors of nuclear magnetic resonance (NMR) signals, particularly from very small samples. For example, nitrogen–vacancy centres in diamond have been used to record NMR signals from nanometre-scale samples, with sensitivity sufficient to detect the magnetic field produced by a single protein. However, the best reported spectral resolution for NMR of molecules using nitrogen–vacancy centres is about 100 hertz. This is insufficient to resolve the key spectral identifiers of molecular structure that are critical to NMR applications in chemistry, structural biology and materials research, such as scalar couplings (which require a resolution of less than ten hertz) and small chemical shifts (which require a resolution of around one part per million of the nuclear Larmor frequency). Conventional, inductively detected NMR can provide the necessary high spectral resolution, but its limited sensitivity typically requires millimetre-scale samples, precluding applications that involve smaller samples, such as picolitre-volume chemical analysis or correlated optical and NMR microscopy. Here we demonstrate a measurement technique that uses a solid-state spin sensor (a magnetometer) consisting of an ensemble of nitrogen–vacancy centres in combination with a narrowband synchronized readout protocol to obtain NMR spectral resolution of about one hertz. We use this technique to observe NMR scalar couplings in a micrometre-scale sample volume of approximately ten picolitres. We also use the ensemble of nitrogen–vacancy centres to apply NMR to thermally polarized nuclear spins and resolve chemical-shift spectra from small molecules. Our technique enables analytical NMR spectroscopy at the scale of single cells.

Space Nuclear Thermal Propulsion Test Facilities Subpanel

NASA Technical Reports Server (NTRS)

Allen, George C.; Warren, John W.; Martinell, John; Clark, John S.; Perkins, David

1993-01-01

On 20 Jul. 1989, in commemoration of the 20th anniversary of the Apollo 11 lunar landing, President George Bush proclaimed his vision for manned space exploration. He stated, 'First for the coming decade, for the 1990's, Space Station Freedom, the next critical step in our space endeavors. And next, for the new century, back to the Moon. Back to the future. And this time, back to stay. And then, a journey into tomorrow, a journey to another planet, a manned mission to Mars.' On 2 Nov. 1989, the President approved a national space policy reaffirming the long range goal of the civil space program: to 'expand human presence and activity beyond Earth orbit into the solar system.' And on 11 May 1990, he specified the goal of landing Astronauts on Mars by 2019, the 50th anniversary of man's first steps on the Moon. To safely and ever permanently venture beyond near Earth environment as charged by the President, mankind must bring to bear extensive new technologies. These include heavy lift launch capability from Earth to low-Earth orbit, automated space rendezvous and docking of large masses, zero gravity countermeasures, and closed loop life support systems. One technology enhancing, and perhaps enabling, the piloted Mars missions is nuclear propulsion, with great benefits over chemical propulsion. Asserting the potential benefits of nuclear propulsion, NASA has sponsored workshops in Nuclear Electric Propulsion and Nuclear Thermal Propulsion and has initiated a tri-agency planning process to ensure that appropriate resources are engaged to meet this exciting technical challenge. At the core of this planning process, NASA, DOE, and DOD established six Nuclear Propulsion Technical Panels in 1991 to provide groundwork for a possible tri-agency Nuclear Propulsion Program and to address the President's vision by advocating an aggressive program in nuclear propulsion. To this end the Nuclear Electric Propulsion Technology Panel has focused it energies; this final report summarizes its endeavor and conclusions.

Thermal interaction of the core and the mantle and long-term behavior of the geomagnetic field

NASA Technical Reports Server (NTRS)

Jones, G. M.

1977-01-01

The effects of temperature changes at the earth's core-mantle boundary on the velocity field of the core are analyzed. It is assumed that the geomagnetic field is maintained by thermal convection in the outer core. A model for the thermal interaction of the core and the mantle is presented which is consistent with current views on the presence of heat sources in the core and the properties of the lower mantle. Significant long-term variations in the frequency of geomagnetic reversals may be the result of fluctuating temperatures at the core-mantle boundary, caused by intermittent convection in the lower mantle. The thermal structure of the lower mantle region D double prime, extending from 2700 to 2900 km in depth, constitutes an important test of this hypothesis and offers a means of deciding whether the geomagnetic dynamo is thermally driven.

Study on preparation and microwave absorption property of the core-nanoshell composite materials doped with La.

PubMed

Wei, Liqiu; Che, Ruxin; Jiang, Yijun; Yu, Bing

2013-12-01

Microwave absorbing material plays a great role in electromagnetic pollution controlling, electromagnetic interference shielding and stealth technology, etc. The core-nanoshell composite materials doped with La were prepared by a solid-state reaction method, which is applied to the electromagnetic wave absorption. The core is magnetic fly-ash hollow cenosphere, and the shell is the nanosized ferrite doped with La. The thermal decomposition process of the sample was investigated by thermogravimetry and differential thermal analysis. The morphology and components of the composite materials were investigated by the X-ray diffraction analysis, the microstructure was observed by scanning electron microscope and transmission electron microscope. The results of vibrating sample magnetometer analysis indicated that the exchange-coupling interaction happens between ferrite of magnetic fly-ash hollow cenosphere and nanosized ferrite coating, which caused outstanding magnetic properties. The microwave absorbing property of the sample was measured by reflectivity far field radar cross section of radar microwave absorbing material with vector network analyzer. The results indicated that the exchange-coupling interaction enhanced magnetic loss of composite materials. Therefore, in the frequency of 5 GHz, the reflection coefficient can achieve -24 dB. It is better than single material and is consistent with requirements of the microwave absorbing material at the low-frequency absorption. Copyright © 2013 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

Characterization of racemic species of chiral drugs using thermal analysis, thermodynamic calculation, and structural studies.

PubMed

Li, Z J; Zell, M T; Munson, E J; Grant, D J

1999-03-01

The identification of the racemic species, as a racemic compound, a racemic conglomerate, or a racemic solid solution (pseudoracemate), is crucial for rationalizing the potential for resolution of racemates by crystallization. The melting points and enthalpies of fusion of a number of chiral drugs and their salts were measured by differential scanning calorimetry. Based on a thermodynamic cycle involving the solid and liquid phases of the enantiomers and racemic species, the enthalpy, entropy and Gibbs free energy of the racemic species were derived from the thermal data. The Gibbs free energy of formation, is always negative for a racemic compound, if it can exist, and the contribution from the entropy of mixing in the liquid state to the free energy of formation is the driving force for the process. For a racemic conglomerate, the entropy of mixing in the liquid state is close to the ideal value of R ln 2 (1.38 cal.mol-1. K-1). Pseudoracemates behave differently from the other two types of racemic species. When the melting points of the racemic species is about 30 K below that of the homochiral species, is approximately zero, indicating that the racemic compound and racemic conglomerate possess similar relative stabilities. The powder X-ray diffraction patterns and 13C solid-state nuclear magnetic resonance spectra are valuable for revealing structural differences between a racemic compound and a racemic conglomerate. Thermodynamic prediction, thermal analysis, and structural study are in excellent agreement for identifying the nature of the racemic species.

The feasibility of thermal and compositional convection in Earth's inner core

NASA Astrophysics Data System (ADS)

Lythgoe, Karen H.; Rudge, John F.; Neufeld, Jerome A.; Deuss, Arwen

2015-05-01

Inner core convection, and the corresponding variations in grain size and alignment, has been proposed to explain the complex seismic structure of the inner core, including its anisotropy, lateral variations and the F-layer at the base of the outer core. We develop a parametrized convection model to investigate the possibility of convection in the inner core, focusing on the dominance of the plume mode of convection versus the translation mode. We investigate thermal and compositional convection separately so as to study the end-members of the system. In the thermal case the dominant mode of convection is strongly dependent on the viscosity of the inner core, the magnitude of which is poorly constrained. Furthermore recent estimates of a large core thermal conductivity result in stable thermal stratification, hindering convection. However, an unstable density stratification may arise due to the pressure dependant partition coefficient of certain light elements. We show that this unstable stratification leads to compositionally driven convection, and that inner core translation is likely to be the dominant convective mode due to the low compositional diffusivity. The style of convection resulting from a combination of both thermal and compositional effects is not easy to understand. For reasonable parameter estimates, the stabilizing thermal buoyancy is greater than the destabilizing compositional buoyancy. However we anticipate complex double diffusive processes to occur given the very different thermal and compositional diffusivities.

The Feasibility of Thermal and Compositional Convection in Earth's Inner Core

NASA Astrophysics Data System (ADS)

Lythgoe, K.; Rudge, J. F.; Neufeld, J. A.; Deuss, A. F.

2014-12-01

Inner core convection, and the corresponding variations in grain size and alignment, has been proposed to explain the complex seismic structure of the inner core, including its anisotropy, lateral variations and the F-layer at the base of the outer core. We develop a parameterised convection model to investigate the possibility of convection in the inner core, focusing on the dominance of the plume mode of convection versus the translation mode. We investigate thermal and compositional convection separately so as to study the end-members of the system. In the thermal case the dominant mode of convection is strongly dependent on the viscosity of the inner core, the magnitude of which is poorly constrained. Furthermore recent estimates of a large core thermal conductivity result in stable thermal stratification, hindering convection. However, an unstable density stratification may arise due to the pressure dependant partition coefficient of certain light elements. We show that this unstable stratification leads to compositionally driven convection, and that inner core translation is likely to be the dominant convective mode due to the low compositional diffusivity. The style of convection resulting from a combination of both thermal and compositional effects is not easy to understand. The stabilising thermal buoyancy is greater than the destabilising compositional buoyancy, however we anticipate complex double diffusive processes to occur given the very different thermal and compositional diffusivities and more work is needed to understand these processes.

Construction and 13C NMR signal-amplification efficiency of a dynamic nuclear polarizer at 6.4 T and 1.4 K

NASA Astrophysics Data System (ADS)

Kiswandhi, Andhika; Niedbalski, Peter; Parish, Christopher; Ferguson, Sarah; Taylor, David; McDonald, George; Lumata, Lloyd

Dissolution dynamic nuclear polarization (DNP) is a rapidly emerging technique in biomedical and metabolic imaging since it amplifies the liquid-state nuclear magnetic resonance (NMR) and imaging (MRI) signals by >10,000-fold. Originally used in nuclear scattering experiments, DNP works by creating a non-Boltzmann nuclear spin distribution by transferring the high electron (γ = 28,000 MHz/T) thermal polarization to the nuclear spins via microwave irradiation of the sample at high magnetic field and low temperature. A dissolution device is used to rapidly dissolve the frozen sample and consequently produces an injectable ``hyperpolarized'' liquid at physiologically-tolerable temperature. Here we report the construction and performance evaluation of a dissolution DNP hyperpolarizer at 6.4 T and 1.4 K using a continuous-flow cryostat. The solid and liquid-state 13C NMR signal enhancement levels of 13C acetate samples doped with trityl OX063 and 4-oxo-TEMPO free radicals will be discussed and compared with the results from the 3.35 T commercial hyperpolarizer. This work is supported by US Dept of Defense Award No. W81XWH-14-1-0048 and Robert A. Welch Foundation Grant No. AT-1877.

Highly Luminescent Zn(x)Cd(1-x)Se/C Core/Shell Nanocrystals: Large Scale Synthesis, Structural and Cathodoluminescence Studies.

PubMed

Bhattacharyya, Sayan; Estrin, Yevgeni; Moshe, Ofer; Rich, Daniel H; Solovyov, Leonid A; Gedanken, A

2009-07-28

Zn(x)Cd(1-x)Se/C core/shell nanocrystals with 31-39 nm semiconducting core and 11-25 nm carbon shell were synthesized from solid state precursors in large scale amounts. A mixture of spherical and tripod nanostructures were obtained only in the one-step reaction (ZC3), where the Zn- and Cd-precursors were reacted simultaneously, rather than in the two step reactions (ZC1 and ZC2), where largely spherical nanostructures were observed. Rietveld analysis of the X-ray diffraction patterns of the samples prepared in three different ways, all under their autogenic pressure, reveal varying compositions of the Zn(x)Cd(1-x)Se nanocrystal core, where the cubic phases with higher Zn content were dominant compared to the hexagonal phases. Carbon encapsulation offers excellent protection to the nanocrystal core and is an added advantage for biological applications. Cathodoluminescence (CL) measurements with spatially integrated and highly localized excitations show distinct peaks and sharp lines at various wavelengths, representing emissions from single nanostructures possessing different compositions, phases, and sizes. Transmission electron microscopy (TEM) showed striations in the nanocrystals that are indicative of a composition modulation, and possibly reveal a phase separation and spinodal decomposition within the nanocrystals. Thermal quenching of the luminescence for both the near band-edge and defect related emissions were observed in the range 60-300 K. The measured activation energies of ∼50-70 meV were related to the presence of shallow donors or acceptors, deep level emissions, and thermal activation and quenching of the luminescence due to the thermal release of electrons from shallow donors to the conduction band or a thermal release of holes from shallow acceptors to the valence band. Spatially integrated CL spectra revealed the existence of broadening and additional components that are consistent with the presence of a composition modulation in the nanocrystals. Spatial localization of the emission in isolated single nanocrystals was studied using monochromatic CL imaging and local CL spectroscopy. CL spectra acquired by a highly localized excitation of individual nanocrystals showed energy shifts in the excitonic luminescence that are consistent with a phase separation into Zn- and Cd-rich regions. The simultaneous appearance of both structural and compositional phase separation for the synthesis of Zn(x)Cd(1-x)Se nanocrystals reveals the complexity and uniqueness of these results.

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Interfacial Ca2+ environments in nanocrystalline apatites revealed by dynamic nuclear polarization enhanced 43Ca NMR spectroscopy

NASA Astrophysics Data System (ADS)

Lee, Daniel; Leroy, César; Crevant, Charlène; Bonhomme-Coury, Laure; Babonneau, Florence; Laurencin, Danielle; Bonhomme, Christian; de Paëpe, Gaël

2017-01-01

The interfaces within bones, teeth and other hybrid biomaterials are of paramount importance but remain particularly difficult to characterize at the molecular level because both sensitive and selective techniques are mandatory. Here, it is demonstrated that unprecedented insights into calcium environments, for example the differentiation of surface and core species of hydroxyapatite nanoparticles, can be obtained using solid-state NMR, when combined with dynamic nuclear polarization. Although calcium represents an ideal NMR target here (and de facto for a large variety of calcium-derived materials), its stable NMR-active isotope, calcium-43, is a highly unreceptive probe. Using the sensitivity gains from dynamic nuclear polarization, not only could calcium-43 NMR spectra be obtained easily, but natural isotopic abundance 2D correlation experiments could be recorded for calcium-43 in short experimental time. This opens perspectives for the detailed study of interfaces in nanostructured materials of the highest biological interest as well as calcium-based nanosystems in general.

A platonic solid templating Archimedean solid: an unprecedented nanometre-sized Ag37 cluster

NASA Astrophysics Data System (ADS)

Li, Xiao-Yu; Su, Hai-Feng; Yu, Kai; Tan, Yuan-Zhi; Wang, Xing-Po; Zhao, Ya-Qin; Sun, Di; Zheng, Lan-Sun

2015-04-01

The spontaneous formation of discrete spherical nanosized molecules is prevalent in nature, but the authentic structural mimicry of such highly symmetric polyhedra from edge sharing of regular polygons has remained elusive. Here we present a novel ball-shaped {(HNEt3)[Ag37S4(SC6H4tBu)24(CF3COO)6(H2O)12]} cluster (1) that is assembled via a one-pot process from polymeric {(HNEt3)2[Ag10(SC6H4tBu)12]}n and CF3COOAg. Single crystal X-ray analysis confirmed that 1 is a Td symmetric spherical molecule with a [Ag36(SC6H4tBu)24] anion shell enwrapping a AgS4 tetrahedron. The shell topology of 1 belongs to one of 13 Archimedean solids, a truncated tetrahedron with four edge-shared hexagons and trigons, which are supported by a AgS4 Platonic solid in the core. Interestingly, the cluster emits green luminescence centered at 515 nm at room temperature. Our investigations have provided a promising synthetic protocol for a high-nuclearity silver cluster based on underlying geometrical principles.The spontaneous formation of discrete spherical nanosized molecules is prevalent in nature, but the authentic structural mimicry of such highly symmetric polyhedra from edge sharing of regular polygons has remained elusive. Here we present a novel ball-shaped {(HNEt3)[Ag37S4(SC6H4tBu)24(CF3COO)6(H2O)12]} cluster (1) that is assembled via a one-pot process from polymeric {(HNEt3)2[Ag10(SC6H4tBu)12]}n and CF3COOAg. Single crystal X-ray analysis confirmed that 1 is a Td symmetric spherical molecule with a [Ag36(SC6H4tBu)24] anion shell enwrapping a AgS4 tetrahedron. The shell topology of 1 belongs to one of 13 Archimedean solids, a truncated tetrahedron with four edge-shared hexagons and trigons, which are supported by a AgS4 Platonic solid in the core. Interestingly, the cluster emits green luminescence centered at 515 nm at room temperature. Our investigations have provided a promising synthetic protocol for a high-nuclearity silver cluster based on underlying geometrical principles. Electronic supplementary information (ESI) available: detailed synthesis procedure, tables, crystal data in CIF files, IR data, TGA results and powder X-ray diffractogram for 1. CCDC 1042228. See DOI: 10.1039/c5nr01222h

Optimization and design of pigments for heat-insulating coatings

NASA Astrophysics Data System (ADS)

Wang, Guang-Hai; Zhang, Yue

2010-12-01

This paper reports that heat insulating property of infrared reflective coatings is obtained through the use of pigments which diffuse near-infrared thermal radiation. Suitable structure and size distribution of pigments would attain maximum diffuse infrared radiation and reduce the pigment volume concentration required. The optimum structure and size range of pigments for reflective infrared coatings are studied by using Kubelka—Munk theory, Mie model and independent scattering approximation. Taking titania particle as the pigment embedded in an inorganic coating, the computational results show that core-shell particles present excellent scattering ability, more so than solid and hollow spherical particles. The optimum radius range of core-shell particles is around 0.3 ~ 1.6 μm. Furthermore, the influence of shell thickness on optical parameters of the coating is also obvious and the optimal thickness of shell is 100-300 nm.

Evaluating Dimethyldiethoxysilane for use in Polyurethane Crosslinked Silica Aerogels

NASA Technical Reports Server (NTRS)

Randall, Jason P.; Meador, Mary Ann B.; Jana, Sadhan C.

2008-01-01

Silica aerogels are highly porous materials which exhibit exceptionally low density and thermal conductivity. Their "pearl necklace" nanostructure, however, is inherently weak; most silica aerogels are brittle and fragile. The strength of aerogels can be improved by employing an additional crosslinking step using isocyanates. In this work, dimethyldiethoxysilane (DMDES) is evaluated for use in the silane backbone of polyurethane crosslinked aerogels. Approximately half of the resulting aerogels exhibited a core/shell morphology of hard crosslinked aerogel surrounding a softer, uncrosslinked center. Solid state NMR and scanning electron microscopy results indicate the DMDES incorporated itself as a conformal coating around the outside of the secondary silica particles, in much the same manner as isocyanate crosslinking. Response surface curves were generated from compression data, indicating levels of reinforcement comparable to that in previous literature, despite the core/shell morphology.

VARIABLE AREA CONTROL ROD FOR NUCLEAR REACTOR

DOEpatents

Huston, N.E.

1960-05-01

A control rod is described which permits continual variation of its absorbing strength uniformly along the length of the rod. The rod is fail safe and is fully inserted into the core but changes in its absorbing strength do not produce axial flux distortion. The control device comprises a sheet containing a material having a high thermal-neutron absorption cross section. A pair of shafts engage the sheet along the longitudinal axis of the shafts and gears associated with the shafts permit winding and unwinding of the sheet around the shafts.

EXPERIMENTAL LIQUID METAL FUEL REACTOR

DOEpatents

Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

1962-01-23

A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Gomez, C. F.; Mireles, O. R.; Stewart, E.

2016-01-01

The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

Preparation of Geophysical Fluid Flow Experiments ( GeoFlow ) in the Fluid Science Laboratory on ISS

NASA Astrophysics Data System (ADS)

Egbers, C.

The ,,GeoFlow" is an ESA experiment planned for the Fluid Science Laboratory on ISS under the scientific coordination (PI) of the Department of Aerodynamics and Fluidmechanics (LAS) at the Brandenburg Technical University (BTU) of Cottbus, Germany. The objective of the experiment is to study thermal convection in the gap between two concentric rotating (full) spheres. A central symmetric force field similar to the gravity field acting on planets can be produced by applying a high voltage between inner and outer sphere using the dielectrophoretic effect (rotating capacitor). To counter the unidirectional gravity under terrestrial conditions, this experiment requires a microgravity environment. The parameters of the experiment are chosen in analogy to the thermal convective motions in the outer core of the Earth. In analogy to geophysical motions in the Earth's liquid core the exp eriment can rotate as solid body as well as differential (inner to outer). Thermal convection is produced by heating the inner sphere and cooling the outer ones. Furtheron, the variation of radius ratio between inner and outer sphere is foreseen as a parameter variation. The flows to be investigated will strongly depend on the gap width and on the Prandtl number. Results of preparatory experiments and numerical simulation of the space experiment will be presented. Funding from DLR under grant 50 WM 0122 is greatfully ackwnoledged.

Core filaments of the nuclear matrix

PubMed Central

1990-01-01

The nuclear matrix is concealed by a much larger mass of chromatin, which can be removed selectively by digesting nuclei with DNase I followed by elution of chromatin with 0.25 M ammonium sulfate. This mild procedure removes chromatin almost completely and preserves nuclear matrix morphology. The complete nuclear matrix consists of a nuclear lamina with an interior matrix composed of thick, polymorphic fibers and large masses that resemble remnant nucleoli. Further extraction of the nuclear matrices of HeLa or MCF-7 cells with 2 M sodium chloride uncovered a network of core filaments. A few dark masses remained enmeshed in the filament network and may be remnants of the nuclear matrix thick fibers and nucleoli. The highly branched core filaments had diameters of 9 and 13 nm measured relative to the intermediate filaments. They may serve as the core structure around which the matrix is constructed. The core filaments retained 70% of nuclear RNA. This RNA consisted both of ribosomal RNA precursors and of very high molecular weight hnRNA with a modal size of 20 kb. Treatment with RNase A removed the core filaments. When 2 M sodium chloride was used directly to remove chromatin after DNase I digestion without a preceding 0.25 M ammonium sulfate extraction, the core filaments were not revealed. Instead, the nuclear interior was filled with amorphous masses that may cover the filaments. This reflected a requirement for a stepwise increase in ionic strength because gradual addition of sodium chloride to a final concentration of 2 M without an 0.25 M ammonium sulfate extraction uncovered core filaments. PMID:2307700

Mars Mission Analysis Trades Based on Legacy and Future Nuclear Propulsion Options

NASA Astrophysics Data System (ADS)

Joyner, Russell; Lentati, Andrea; Cichon, Jaclyn

2007-01-01

The purpose of this paper is to discuss the results of mission-based system trades when using a nuclear thermal propulsion (NTP) system for Solar System exploration. The results are based on comparing reactor designs that use a ceramic-metallic (CERMET), graphite matrix, graphite composite matrix, or carbide matrix fuel element designs. The composite graphite matrix and CERMET designs have been examined for providing power as well as propulsion. Approaches to the design of the NTP to be discussed will include an examination of graphite, composite, carbide, and CERMET core designs and the attributes of each in regards to performance and power generation capability. The focus is on NTP approaches based on tested fuel materials within a prismatic fuel form per the Argonne National Laboratory testing and the ROVER/NERVA program. NTP concepts have been examined for several years at Pratt & Whitney Rocketdyne for use as the primary propulsion for human missions beyond earth. Recently, an approach was taken to examine the design trades between specific NTP concepts; NERVA-based (UC)C-Graphite, (UC,ZrC)C-Composite, (U,Zr)C-Solid Carbide and UO2-W CERMET. Using Pratt & Whitney Rocketdyne's multidisciplinary design analysis capability, a detailed mission and vehicle model has been used to examine how several of these NTP designs impact a human Mars mission. Trends for the propulsion system mass as a function of power level (i.e. thrust size) for the graphite-carbide and CERMET designs were established and correlated against data created over the past forty years. These were used for the mission trade study. The resulting mission trades presented in this paper used a comprehensive modeling approach that captures the mission, vehicle subsystems, and NTP sizing.

Stopping power beyond the adiabatic approximation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caro, M.; Correa, A. A.; Artacho, E.

2017-06-01

Energetic ions traveling in solids deposit energy in a variety of ways, being nuclear and electronic stopping the two avenues in which dissipation is usually treated. This separation between electrons and ions relies on the adiabatic approximation in which ions interact via forces derived from the instantaneous electronic ground state. In a more detailed view, in which non-adiabatic effects are explicitly considered, electronic excitations alter the atomic bonding, which translates into changes in the interatomic forces. In this work, we use time dependent density functional theory and forces derived from the equations of Ehrenfest dynamics that depend instantaneously on themore » time-dependent electronic density. With them we analyze how the inter-ionic forces are affected by electronic excitations in a model of a Ni projectile interacting with a Ni target, a metallic system with strong electronic stopping and shallow core level states. We find that the electronic excitations induce substantial modifications to the inter-ionic forces, which translate into nuclear stopping power well above the adiabatic prediction. Particularly, we observe that most of the alteration of the adiabatic potential in early times comes from the ionization of the core levels of the target ions, not readily screened by the valence electrons.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

Liquid-Solid Interaction in Al-Si/Al-Mn-Cu-Mg Brazing Sheets and Its Effects on Mechanical Properties

NASA Astrophysics Data System (ADS)

Jin, H.; Kozdras, M. S.; Amirkhiz, B. Shalchi; Winkler, S. L.

2018-05-01

The liquid-solid interaction during brazing at 592 °C to 605 °C and its effects on mechanical properties were investigated in a series of Al-Si/Al-Mn-Cu-Mg brazing sheets with different Mg contents. Depending on the Mg level in core alloy and the brazing temperature, critical changes of local chemistry and microstructure related to the liquid-solid interaction occur, including solid-state diffusion, uniform clad-core interface migration, and grain boundary penetration (GBP). When the Mg in core alloy is below 1 wt pct, the interaction is limited and the formation of a dense precipitation band due to solid-state diffusion of Si from the clad to the core is dominant. As the Mg exceeds 1 wt pct, very extensive interaction occurs resulting in clad-core interface migration and GBP of Si into the core, both involving local melting and re-solidification of the core alloy. Whenever Si from the clad encounters Mg in the core due to the interaction, Mg2Si precipitates are formed leading to significant improvement of strength. However, the interface migration and GBP drastically reduce the ductility, due to the segregation of coarse secondary phase particles along the newly formed grain boundaries.

Liquid-Solid Interaction in Al-Si/Al-Mn-Cu-Mg Brazing Sheets and Its Effects on Mechanical Properties

NASA Astrophysics Data System (ADS)

Jin, H.; Kozdras, M. S.; Amirkhiz, B. Shalchi; Winkler, S. L.

2018-07-01

The liquid-solid interaction during brazing at 592 °C to 605 °C and its effects on mechanical properties were investigated in a series of Al-Si/Al-Mn-Cu-Mg brazing sheets with different Mg contents. Depending on the Mg level in core alloy and the brazing temperature, critical changes of local chemistry and microstructure related to the liquid-solid interaction occur, including solid-state diffusion, uniform clad-core interface migration, and grain boundary penetration (GBP). When the Mg in core alloy is below 1 wt pct, the interaction is limited and the formation of a dense precipitation band due to solid-state diffusion of Si from the clad to the core is dominant. As the Mg exceeds 1 wt pct, very extensive interaction occurs resulting in clad-core interface migration and GBP of Si into the core, both involving local melting and re-solidification of the core alloy. Whenever Si from the clad encounters Mg in the core due to the interaction, Mg2Si precipitates are formed leading to significant improvement of strength. However, the interface migration and GBP drastically reduce the ductility, due to the segregation of coarse secondary phase particles along the newly formed grain boundaries.

Key characteristics of the Fe-snow regime in Ganymede's core

NASA Astrophysics Data System (ADS)

Rückriemen, Tina; Breuer, Doris; Spohn, Tilman

2014-05-01

Ganymede shows signs of an internally produced dipolar magnetic field (|Bdip|≡719 nT) [1]. For small planetary bodies such as Ganymede the Fe-snow regime, i.e. the top-down solidification of iron, has been suggested to play an important role in the core cooling history [2,3]. In that regime, iron crystals form first at the core-mantle boundary (CMB) due to shallow or negative slopes of the melting temperature [2,3]. The solid iron particles are heavier than the surrounding Fe-FeS fluid, i.e. a snow zone forms, settle to deeper core regions, where the core temperature is higher than the melting temperature, and remelt again. As a consequence, a stable chemical gradient in the Fe-FeS fluid arises within the snow zone. We speculate this style of convection via sedimentation to be small scale, therefore it lacks an important criterion necessary for dynamo action [4]. Below this zone, whose thickness increases with time, the process of remelting of iron creates a gravitationally unstable situation. We propose that this could be the driving mechanism for a potential dynamo. However, dynamo action would be restricted to the time period the snow zone needs to grow across the core. With a 1D thermo-chemical evolution model, we investigate key characteristics of the Fe-snow regime within Ganymede's core: the compositional density gradient of the fluid Fe-FeS within the snow zone and the time period necessary to grow the snow zone across the core. Additionally, we determine the dipolar magnetic field strength associated with a dynamo in Ganymede's deeper fluid core. We vary important input paramters such as the initial sulfur concentration (7-19 wt.%), the core heat flux (2-6 mW/m2) and the thermal conductivity (20-60 W/mK) with the nominal model being: xs=10 wt.%, qcmb=4 mW/m2, kc=32 W/mK. We find, that heat fluxes higher than 6 or 22 mW/m2 are required for double-diffusive or overturning convection to overcome the compositional density gradient within the snow zone, respectively. Since Ganymede's core heat flux does not exceed values of 4 mW/m2 [2], we consider the snow zone to be stable against thermal convection. The time necessary to grow the snow zone across the core is between 230-1900 Myr. For representative models we calculate the temporal evolution of the surface dipolar magnetic field strength according to [5]. All models show surface dipolar magnetic field strengths during the evolution of the snow zone that match the observed value of |Bdip|≡719 nT. In conclusion, we find that the Fe-snow regime produces a stably-stratified liquid layer in the snow zone below which a magnetic field of observed strength can be generated. Such a chemical dynamo is restricted in time and stops as soon as an inner solid core starts to grow suggesting the absence of such an inner core in Ganymede. The present model further suggests a core with high initial sulfur concentration, because this leads to a late start and a long duration of the dynamo necessary to explain the present magnetic field. References [1] Kivelson, M et al. (1996), Nature, 384(6609), [2] Hauck II, S. et al. (2006), JGR, 111(E9), [3] Williams, Q. (2009), EPSL, 284(3), [4] Christensen, U. and J. Wicht (2007), Treatise of Geophysics, Elsevier, [5] Christensen, U., and J. Aubert (2006), GJI, 166(1)

Thermal Diffusion Fractionation of Cr and V Isotope in Silicate Melt

NASA Astrophysics Data System (ADS)

Lin, X.; Lundstrom, C.

2017-12-01

Earth's mantle is isotopically heavy relative to chondrites for V, Cr and some other siderophile elements. A possible solution is that isotopic fractionation by thermal diffusion occurs in a thermal boundary layer between solid mantle and an underlying basal magma ocean (BMO:Labrosse et al.,2007). If so, isotopically light composition might partition into the core, resulting in a complimentary isotopically heavy solid mantle. To verify how much fractionation could happen in this process, piston cylinder experiment were conducted to investigate the fractionation of Cr and V isotope ratios in partially molten silicate under an imposed temperature gradient from 1650 °C to 1350 °C at 1 GPa for 10 to 50 hours to reach a steady state isotopic profile. The temperature profile for experiments was determined by the spinel-growth method at the same pressure and temperature. Experimental runs result in 100% glass at the hot end progressing to nearly 100 % olivine at the cold end. Major and minor element concentrations of run products show systematic changes with temperature. Glass MgO contents increase and Al2O3 and CaO contents decrease by several weight percent as temperature increases across the charge. These are well modeled using IRIDIUM (Boudreau 2003) to simulate the experiments. Isotopic composition measurements of Cr and V at different temperatures are in progress, providing the first determinations of thermal diffusion isotopic sensitivity, Ω (permil isotopic fractionation per temperature offset per mass unit) for these elements. These results will be compared with previously determined Ω for network formers and modifiers and used in a BMO-based thermal diffusion model for formation of Earth's isotopically heavy mantle.

Theory of the spin-1 bosonic liquid metal - Equilibrium properties of liquid metallic deuterium

NASA Technical Reports Server (NTRS)

Oliva, J.; Ashcroft, N. W.

1984-01-01

The theory of a two-component quantum fluid comprised of spin-1/2 fermions and nonzero spin bosons is examined. This system is of interest because it embodies a possible quantum liquid metallic phase of highly compressed deuterium. Bose condensation is assumed present and the two cases of nuclear-spin-polarized and -unpolarized systems are considered. A significant feature in the unpolarized case is the presence of a nonmagnetic mode with quadratic dispersion owing its existence to nonzero boson spin. The physical character of this mode is examined in detail within a Bogoliubov approach. The specific heat, bulk modulus, spin susceptibility, and thermal expansion are all determined. Striking contrasts in the specific heats and thermal-expansion coefficients of the liquid and corresponding normal solid metallic phase are predicted.

Thermal reactions of uranium metal, UO 2, U 3O 8, UF 4, and UO 2F 2 with NF 3 to produce UF 6

NASA Astrophysics Data System (ADS)

McNamara, Bruce; Scheele, Randall; Kozelisky, Anne; Edwards, Matthew

2009-11-01

This paper demonstrates that NF 3 fluorinates uranium metal, UO 2, UF 4, UO 3, U 3O 8, and UO 2F 2·2H 2O to produce the volatile UF 6 at temperatures between 100 and 550 °C. Thermogravimetric and differential thermal analysis reaction profiles are described that reflect changes in the uranium fluorination/oxidation state, physiochemical effects, and instances of discrete chemical speciation. Large differences in the onset temperatures for each system investigated implicate changes in mode of the NF 3 gas-solid surface interaction. These studies also demonstrate that NF 3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in actinide volatility reprocessing.

Advanced Materials and Solids Analysis Research Core (AMSARC)

EPA Science Inventory

The Advanced Materials and Solids Analysis Research Core (AMSARC), centered at the U.S. Environmental Protection Agency's (EPA) Andrew W. Breidenbach Environmental Research Center in Cincinnati, Ohio, is the foundation for the Agency's solids and surfaces analysis capabilities. ...

A New Star-shaped Carbazole Derivative with Polyhedral Oligomeric Silsesquioxane Core: Crystal Structure and Unique Photoluminescence Property.

PubMed

Xu, Zixuan; Yu, Tianzhi; Zhao, Yuling; Zhang, Hui; Zhao, Guoyun; Li, Jianfeng; Chai, Lanqin

2016-01-01

A new inorganic–organic hybrid material based on polyhedral oligomeric silsesquioxane (POSS) capped with carbazolyl substituents, octakis[3-(carbazol-9-yl)propyldimethylsiloxy]-silsesquioxane (POSS-8Cz), was successfully synthesized and characterized. The X-ray crystal structure of POSS-8Cz were described. The photophysical properties of POSS-8Cz were investigated by using UV–vis,photoluminescence spectroscopic analysis. The hybrid material exhibits blue emission in the solution and the solid film.The morphology and thermal stablity properties were measured by X-ray diffraction (XRD) and TG-DTA analysis.

Disk-Planet Torques from Radiation-Hydrodynamics Calculations with Spatially-Resolved Planetary Envelopes Undergoing Solids' Accretion

NASA Astrophysics Data System (ADS)

D'Angelo, G.

2016-12-01

D'Angelo & Bodenheimer (2013, ApJ, 778, 77) performed global 3D radiation-hydrodynamics disk-planet simulations aimed at studying envelope formation around planetary cores, during the phase of sustained planetesimal accretion. The calculations modeled cores of 5, 10, and 15 Earth masses orbiting a sun-like star in a protoplanetary disk extending from ap/2 to 2ap in radius, ap=5 or 10 AU being the core's orbital radius. The gas equation of state - for a solar mixture of H2, H, He - accounted for translational, rotational, and vibrational states, for molecular dissociation and atomic ionization, and for radiation energy. Dust opacity calculations applied the Mie theory to multiple grain species whose size distributions ranged from 5e-6 to 1 mm. Mesh refinement via grid nesting allowed the planets' envelopes to be resolved at the core-radius length scale. Passive tracers were used to determine the volume of gas bound to a core, defining the envelope, and resulting in planet radii comparable to the Bondi radius. The energy budjet included contributions from the accretion of solids on the cores, whose rates were self-consistently computed with a 1D planet formation code. At this stage of the planet's growth, gravitational energy released in the envelope by solids' accretion far exceeds that released by gas accretion. These models are used to determine the gravitational torques exerted by the disk's gas on the planet and the resulting orbital migration rates. Since the envelope radius is a direct product of the models, they allow for a non-ambiguous assessment of the torques exerted by gas not bound to the planet. Additionally, since planets' envelopes are fully resolved, thermal and dynamical effects on the surrounding disk's gas are accurately taken into account. The computed migration rates are compared to those obtained from existing semi-analytical formulations for planets orbiting in isothermal and adiabatic disks. Because these formulations do not account for thermodynamical interactions between the planet's envelope and the disk's gas, the numerical models are also used to quanitfy the impact of short-scale tidal interactions on the total torque acting on the planet. Computing resources were provided by the NASA High-End Computing Program through the NASA Advanced Supercomputing Division at Ames Research Center.

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

NEUTRONIC REACTOR CORE

DOEpatents

Thomson, W.B.; Corbin, A. Jr.

1961-07-18

An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

Testing in Support of Space Fission System Development and Qualification

NASA Technical Reports Server (NTRS)

Houts, Mike; Bragg-Sitton, Shannon; Garber, Anne; Godfrey, Tom; Martin, Jim; Pearson, Boise; Webster, Kenny

2007-01-01

Extensive data would be required for the qualification of a fission surface power (FSP) system. The strategy for qualifying a FSP system could have a significant programmatic impact. This paper explores potential options that could be used for qualifying FSP systems, including cost-effective means for obtaining required data. three methods for obtaining qualification data are analysis, non-nuclear testing, and nuclear testing. It has been over 40 years since the US qualified a space reactor for launch. During that time, advances have been made related to all three methods. Perhaps the greatest advancement has occurred in the area of computational tools for design and analysis. Tools that have been developed, coupled with modem computers, would have a significant impact on a FSP qualification. This would be especially true for systems with materials and fuels operating well within temperature, irradiation damage, and burnup limits. The ability to perform highly realistic non-nuclear testing has also advanced throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modem FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.} and extensive data to be taken from the core region. Both steady-state and transient operation can be tested. For transient testing, reactivity feedback is calculated (or measured in cold/warm criticals) based on reactor temperature and/or dimensional changes. Pin power during a transient is then calculated based on the reactivity feedback that would occur given measured values of temperature and/or dimensional change. In this way nonnuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. Realistic non-nuclear testing is most useful for systems operating within known temperature, irradiation damage, and burnup capabilities.

Mechanical and photoelastic analysis of conventional screws and cannulated screws for sagittal split osteotomy fixation: a comparative study.

PubMed

Lima, Cristina Jardelino de; Falci, Saulo Gabriel Moreira; Rodrigues, Danillo Costa; Marchiori, Érica Cristina; Moreira, Roger Willian Fernandes

2015-12-01

The aim of the present study was to use mechanical and photoelastic tests to compare the performance of cannulated screws with solid-core screws in sagittal split osteotomy fixation. Ten polyurethane mandibles, with a prefabricated sagittal split ramus osteotomy, were fixed with an L inverted technique and allocated to each group as follows: cannulated screw group (CSG), fixed with three 2.3-cannulated screws; and solid-core screw group (SCSG), fixed with three 2.3-solid-core screws. Vertical linear loading tests were performed. The differences between mean values were analyzed through T test for independent samples. The photoelastic test was carried out using a polariscope. The results revealed differences between the two groups only at 1 mm of displacement, in which the cannulated-screw revealed more resistance. Photoelastic test showed higher stress concentration close to mandibular branch in the solid-core group. Cannulated screws performed better than solid-core ones in a mechanical test at 1-mm displacement and photoelastic tests.

Modelling and Characterization of Effective Thermal Conductivity of Single Hollow Glass Microsphere and Its Powder.

PubMed

Liu, Bing; Wang, Hui; Qin, Qing-Hua

2018-01-14

Tiny hollow glass microsphere (HGM) can be applied for designing new light-weighted and thermal-insulated composites as high strength core, owing to its hollow structure. However, little work has been found for studying its own overall thermal conductivity independent of any matrix, which generally cannot be measured or evaluated directly. In this study, the overall thermal conductivity of HGM is investigated experimentally and numerically. The experimental investigation of thermal conductivity of HGM powder is performed by the transient plane source (TPS) technique to provide a reference to numerical results, which are obtained by a developed three-dimensional two-step hierarchical computational method. In the present method, three heterogeneous HGM stacking elements representing different distributions of HGMs in the powder are assumed. Each stacking element and its equivalent homogeneous solid counterpart are, respectively, embedded into a fictitious matrix material as fillers to form two equivalent composite systems at different levels, and then the overall thermal conductivity of each stacking element can be numerically determined through the equivalence of the two systems. The comparison of experimental and computational results indicates the present computational modeling can be used for effectively predicting the overall thermal conductivity of single HGM and its powder in a flexible way. Besides, it is necessary to note that the influence of thermal interfacial resistance cannot be removed from the experimental results in the TPS measurement.

Solid state phase change materials for thermal energy storage in passive solar heated buildings

NASA Astrophysics Data System (ADS)

Benson, D. K.; Christensen, C.

1983-11-01

A set of solid state phase change materials was evaluated for possible use in passive solar thermal energy storage systems. The most promising materials are organic solid solutions of pentaerythritol, pentaglycerine and neopentyl glycol. Solid solution mixtures of these compounds can be tailored so that they exhibit solid-to-solid phase transformations at any desired temperature within the range from less than 25 deg to 188 deg. Thermophysical properties such as thermal conductivity, density and volumetric expansion were measured. Computer simulations were used to predict the performance of various Trombe wall designs incorporating solid state phase change materials. Optimum performance was found to be sensitive to the choice of phase change temperatures and to the thermal conductivity of the phase change material. A molecular mechanism of the solid state phase transition is proposed and supported by infrared spectroscopic evidence.

Comprehensive study of thermal properties of lunar core samples

NASA Technical Reports Server (NTRS)

Langseth, M. G.; Horath, K.

1975-01-01

The feasibility of a technique for measuring the thermal conductivity of lunar core samples was investigated. The thermal conduction equation for a composite cylinder was solved to obtain a mathematical expression for the surface temperature of the core tube filled with lunar material. The sample is heated by radiation from the outside at a known rate, the variation of the temperature at the surface of the core tube is measured, and the thermal conductivity determined by comparing the observed temperature with the theoretically expected one. The apparatus used in the experiment is described.

Turbine component casting core with high resolution region

DOEpatents

Kamel, Ahmed; Merrill, Gary B.

2014-08-26

A hollow turbine engine component with complex internal features can include a first region and a second, high resolution region. The first region can be defined by a first ceramic core piece formed by any conventional process, such as by injection molding or transfer molding. The second region can be defined by a second ceramic core piece formed separately by a method effective to produce high resolution features, such as tomo lithographic molding. The first core piece and the second core piece can be joined by interlocking engagement that once subjected to an intermediate thermal heat treatment process thermally deform to form a three dimensional interlocking joint between the first and second core pieces by allowing thermal creep to irreversibly interlock the first and second core pieces together such that the joint becomes physically locked together providing joint stability through thermal processing.

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis - Midloop Operation with Loss of Residual Heat Removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulatedmore » by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.« less

Dielectric Heaters for Testing Spacecraft Nuclear Reactors

NASA Technical Reports Server (NTRS)

Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

2006-01-01

A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.