Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “CORE”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Formed Core Sampler Hydraulic Conductivity Testing

A full-scale formed core sampler was designed and functionally tested for use in the Saltstone Disposal Facility (SDF). Savannah River National Laboratory (SRNL) was requested to compare properties of the formed core samples and core drilled samples taken from adjacent areas in the full-scale sampler. While several physical properties were evaluated, the primary property of interest was hydraulic conductivity. Differences in hydraulic conductivity between the samples from the formed core sampler and those representing the bulk material were noted with respect to the initial handling and storage of the samples. Due to testing conditions, the site port samples were exposed to uncontrolled temperature and humidity conditions prior to testing whereas the formed core samples were kept in sealed containers with minimal exposure to an uncontrolled environment prior to testing. Based on the results of the testing, no significant differences in porosity or density were found between the formed core samples and those representing the bulk material in the test stand.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Effect of Carbon on the Volume of Solid Iron at High Pressure: Implications for Carbon Substitution in Iron Structures and Carbon Content in the Earth’s Inner Core

Understanding the effect of carbon on the density of hcp (hexagonal-close-packed) Fe-C alloys is essential for modeling the carbon content in the Earth’s inner core. Previous studies have focused on the equations of state of iron carbides that may not be applicable to the solid inner core that may incorporate carbon as dissolved carbon in metallic iron. Carbon substitution in hcp-Fe and its effect on the density have never been experimentally studied. We investigated the compression behavior of Fe-C alloys with 0.31 and 1.37 wt % carbon, along with pure iron as a reference, by in-situ X-ray diffraction measurements up to 135 GPa for pure Fe, and 87 GPa for Fe-0.31C and 109 GPa for Fe-1.37C. The results show that the incorporation of carbon in hcp-Fe leads to the expansion of the lattice, contrary to the known effect in body-centered cubic (bcc)-Fe, suggesting a change in the substitution mechanism or local environment. The data on axial compressibility suggest that increasing carbon content could enhance seismic anisotropy in the Earth’s inner core. The new thermoelastic parameters allow us to develop a thermoelastic model to estimate the carbon content in the inner core when carbon is incorporated as dissolved carbon hcp-Fe. The required carbon contents to explain the density deficit of Earth’s inner core are 1.30 and 0.43 wt % at inner core boundary temperatures of 5000 K and 7000 K, respectively.

58 GEOSCIENCES↗
DOE OSTI2020

C HIMERA : A Massively Parallel Code for Core-collapse Supernova Simulations

Herein we provide a detailed description of the C HIMERA code, a code developed to model core collapse supernovae (CCSNe) in multiple spatial dimensions. The CCSN explosion mechanism remains the subject of intense research. Progress to date demonstrates that it involves a complex interplay of neutrino production, transport, and interaction in the stellar core, three-dimensional stellar core fluid dynamics and its associated instabilities, nuclear burning, and the fundamental physics of the neutrino–stellar core weak interactions and the equations of state of all stellar core constituents—particularly, the nuclear equation of state associated with core nucleons, both free and bound in nuclei. C HIMERA , by incorporating detailed neutrino transport, realistic neutrino–matter interactions, three-dimensional hydrodynamics, realistic nuclear, leptonic, and photonic equations of state, and a nuclear reaction network, along with other refinements, can be used to study the role of neutrino radiation, hydrodynamic instabilities, and a variety of input physics in the explosion mechanism itself. It can also be used to compute observables such as neutrino signatures, gravitational radiation, and the products of nucleosynthesis associated with CCSNe. The code contains modules for neutrino transport, multidimensional compressible hydrodynamics, nuclear reactions, a variety of neutrino interactions, equations of state, and modules to provide data for post-processing observables such as the products of nucleosynthesis, and gravitational radiation. C HIMERA is an evolving code, being updated periodically with improved input physics and numerical refinements. We detail here the current version of the code, from which future improvements will stem, which can in turn be described as needed in future publications.

79 ASTRONOMY AND ASTROPHYSICS↗
NASA NTRS2022

The Lithophile Element Budget of Earth’s Core

The relative composition of Earth’s core and mantle were set during core formation. By determining how elements partition between metal and silicate at high pressures and temperatures, measurements of the mantle composition and geophysical observations of the core can be used to understand the mechanisms by which Earth formed. Here we present the results of metal–silicate partitioning experiments for a range of nominally lithophile elements (Al, Ca, K, Mg, O, S, Si, Th, U) and S to 85 GPa and up to 5400 K. With our results and a compilation of literature data, we developed a parameterization for partitioning that accounts for compositional dependencies in both the metal and silicate phases. Using this parameterization in a range of planetary growth models, we find that, in general, lithophile element partitioning into the metallic phase is enhanced at high temperatures. The relative abundances of FeO, SiO2 and MgO in the mantle vary significantly between planetary growth models, and the mantle abundances of these elements can be used to provide important constraints on Earth’s accretion. To match Earth’s core mass and mantle composition, Earth’s building blocks must have been enriched in Fe and depleted in Si compared with CI chondrites. Finally, too little Mg, Si and O are partitioned into the core for precipitation of oxides to be a major source of energy for the geodynamo. In contrast, several ppb of U can be partitioned into the core at high temperatures, and this energy source must be accounted for in thermal evolution models.

dynamo↗
NASA NTRS2018

Core-Mantle Fractionation of Carbon in Earth and Mars: the Effects of Sulfur

Constraining carbon (C) fractionation between silicate magma ocean (MO) and core-forming alloy liquid during early differentiation is essential to understand the origin and early distribution of C between reservoirs such as the crust-atmosphere, mantle, and core of Earth and other terrestrial planets. Yet experimental data at high pressure (P)-temperature (T) on the effect of other light elements such as sulfur (S) in alloy liquid on alloy-silicate partitioning of C and C solubility in Fe-alloy compositions relevant for core formation is lacking. Here we have performed multi-anvil experiments at 6–13 GPa and 1800–2000 °C to examine the effects of S and Ni on the solubility limit of C in Fe-rich alloy liquid as well as partitioning behavior of C between alloy liquid and silicate melt (𝐷𝐷(sub C)(sup alloy/silicate)). The results show that C solubility in the alloy liquid as well as 𝐷𝐷(sub C)(sup alloy/silicate) decreases with increasing in S content in the alloy liquid. Empirical regression on C solubility in alloy liquid using our new experimental data and previous experiments demonstrates that C solubility significantly increases with increasing temperature, whereas unlike in S-poor or S-free alloy compositions, there is no discernible effect of Ni on C solubility in S-rich alloy liquid. Our modelling results confirm previous findings that in order to satisfy the C budget of BSE, the bulk Earth C undergoing alloy-silicate fractionation needs to be as high as those of CI-type carbonaceous chondrite, i.e., not leaving any room for volatility-induced loss of carbon during accretion. For Mars, on the other hand, an average single-stage core formation at relatively oxidized conditions (1.0 log unit below IW buffer) with 10-16 wt.% S in the core could yield a Martian mantle with a C budget similar to that of Earth’s BSE for a bulk C content of ~0.25-0.9 wt.%. For the scenario where C was delivered to the proto-Earth by a S-rich differentiated impactor at a later stage, our model calculations predict that bulk C content in the impactor can be as low as ~0.5 wt.% for an impactor mass that lies between 9-20% of present day Earth’s mass. This value is much higher than 0.05-0.1 wt.% bulk C in the impactor predicted by Li et al. (2016) because C-solubility limit of 0.3 wt.% in a S-rich alloy predicted by their models is significantly lower than the experimentally derived C-solubility of ~1.6 wt. % for the relevant S-content in the core of the impactor.

Carbon↗
NASA NTRSDate not supplied

Green Run Modal Test of the NASA Space Launch System Core Stage

The Core Stage of the new NASA Space Launch System (SLS) is a 212-foot-tall rocket assembly—consisting primarily of two cryogenic propellant tanks, an engine section, and four RS-25 rocket engines—that will send crew and large payloads to the moon and beyond for NASA’s Artemis program. Prior to SLS assembly, the Core Stage completed a series of structural and functional tests in the B-2 Test Stand at Stennis Space Center, designated Green Run. The goal of Green Run was to verify analytical models, confirm proper subsystem operation, and test-fire all four RS-25 engines of the Core Stage. In January 2020, Green Run testing began with an experimental modal analysis test, performed by the Marshal Space Flight Center modal test team. A free-boundary test condition of the Core Stage was simulated as close as possible by suspending the massive launch vehicle from the B-2 Test Stand crane. Modal excitation was provided by a pair of 250-lb electro-dynamic shakers for multi-shaker random vibration testing, as well as a 12-pound instrumented hammer for impact testing. Modal response was measured with 550 accelerometer channels distributed on both the Core Stage and the B-2 Test Stand derrick crane. Following one very long day of testing, frequency response functions were calculated from the measured time histories in the target mode frequency range of 5 Hz to 15 Hz, and mode shapes, frequencies, and damping values were successfully estimated. The case-study presented in this paper will discuss Green Run, the SLS Core Stage, the modal test setup and execution, as well as a brief overview of the test results. Challenges associated with testing such a large, suspended structure in an outdoor environment will be discussed as well.

Space Launch System↗
DOE OSTI2012

MCNP LWR Core Generator

The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

UT-GOM2-1 Hydrate Pressure Coring Expedition Prospectus

The UT-GOM2-1 Hydrate Pressure Coring Expedition will drill two holes in Green Canyon Block 955 in the deepwater Gulf of Mexico to pressure core and wireline log a coarse-grained methane hydrate reservoir. We will take ten, 10 ft, pressure cores in each well. The cores will be imaged and logged under pressure. Samples will be degassed to interpret hydrate concentration. Microbiological and pore water analyses will be performed. 20 3 ft pressure cores will be returned to shore for analysis. The log and core data will provide a foundation for scientific exploration by the greater hydrate research community

03 NATURAL GAS↗
DOE OSTI2022

Initial study on cross section generation requirements for a PBR equilibrium core

A Serpent model of the HTR-PM equilibrium core was developed for use in cross section prepa-ration studies in order to guide methods development for the Griffin reactor multiphysics applica-tion. The model includes detailed isotopics for 10 distinct pebble burnup groups in 126 core zoneswith unique fuel and moderator temperatures obtained from a coupled neutronics-thermal-fluidsequilibrium core calculation using Griffin-Pronghorn. A sensitivity study of the fuel and mod-erator temperatures for various core regions was performed with the MOOSE stochastic tools.The results show that the uncertainties are, not unexpectedly, dominated by the value of the fluidtemperature and that the power level, heat transfer coefficient and effective conduction to neigh-boring pebbles and fluid constitute, at best, second order effects. The temperature uncertaintyrange varies from 28 K to 57 K at the core entry and exit planes, respectively, but these val-ues are probably higher. We still have to quantify the significance of these uncertainties in thepreparation of cross sections, which will be postponed for future work. In addition, we verifythat the single effective pebble approach works well for the preparation of region averaged crosssections in the infinite domain approximation. Nevertheless, there are significant discrepanciesin the cross sections when compared to the multi-pebble model. This could affect the predictionof peak values and in the depletion calculation. We conclude that is highly desirable for futurestudies with Griffin to be able to handle both the ?effective? pebble approximation and the multi-pebble approach for the various pebble burnup groups. This enables Griffin users the flexibilityto perform higher-fidelity studies. Finally, we initiate the preparation of cross sections for variouscore regions from the full core Serpent reference model. We quantify the differences in 26 groupcross sections from infinite domain models. These reference cross sections will serve to validatethe double heterogeneity, self-shielding, and spectrum-correction methods in Griffin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Development of flow network analysis code for block type VHTR core by linear theory method

VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Contribution to modeling of the reflooding of a severely damaged reactor core using PRELUDE experimental results

In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods may result in the formation of a 'debris bed'. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. The PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Role of Core Exit Axial Reflectors for NTP Full Submersion Criticality Accident Mitigation

One of the only design basis accidents currently codified into law for space nuclear systems is the full-submersion criticality accident. This accident scenario prescribes that the reactor core must remain sub-critical in the event of a full water submersion with the control system in the shut-down position. This accident scenario is of particular interest and relevance for nuclear thermal propulsion (NTP) reactors because of their unique core periphery and internal design considerations. In previous work it has been shown that this accident is particularly sever for low-enriched uranium (LEU) fueled NTP cores due to their moderated spectrum and the minimization of parasitic neutron absorption. These then add to the existing problems associated with the numerous voids within the core and the extensive neutron leakage due to its compact size. In this abstract we examine the role of a core exit axial reflector in minimizing the submersion worth of a graphite composite LEU-NTP core. The neutronics simulations were done using the ENDF/B-VII cross-section library in conjunction with MCNP6.1. The MCNP calculations were done using 20000 particles per cycle with 50 inactive and 300 active cycles. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Small CaH{sub 2} Moderated Thermal Reactor for Surface Power Generation II: Core Neutronics Design

National Aeronautics and Space Administration (NASA) is considering to adopt a small nuclear reactor for the planned Mars surface mission. The required specifications of the reactor system are: 10- to 100-kW electric power with power density above 5 We/kg; 8-year stable operation. In the proposed study, stretch goals for this surface power system were set as: electric power of 20 kWe with total system weight below 1000 kg. Furthermore, nuclear security and nuclear safety in the case of launch accidents must be considered for surface power systems, therefore, selecting low enriched uranium (LEU, = 20 wt% of {sup 235}U/U) is preferable due to its lower criticality. However, the reactor core weight tends to increase in the case of a reactor with LEU fuel, because relatively larger-mass core is required to achieve the critical. The essential feature of the space reactor core neutronics design is to balance between the nuclear security and safety, and the mass reduction of the core. Some feasible methods to reduce the core mass in the case of the LEU core are: (1) optimized neutron moderation, (2) reduce neutron leakage, and (3) increase multiplication of neutron. A solid moderator has been proposed by Kimura et.al.: neutron leakage control and neutron multiplicity will be featured in the current study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Spanning the Core Design Space with Binary Perturbations - A Mathematical Investigation

A reactor core design engineer, even one with many years of experience or one using advanced fuel pattern optimization software, will likely fine tune the pattern before finalizing the design. Fine tuning can further minimize local enthalpy peaks, optimize reactivity to extend burnup, and add additional margin to limits on reactivity coefficients or heat flux. For reactors that have reached an equilibrium fuel cycle, the number of feeds each cycle is typically the similar to previous cycles as is their placement in the core. It is then the engineer's responsibly to efficiently select the placement of burned assemblies to ensure optimal fuel utilization, satisfy safety limits, and meet cycle energy requirements. One of the engineer's most commonly used tools in this fine tuning investigation is to swap the positions of two assemblies, most likely two previously irradiated ('burned') assemblies since their reactivity is lower than fresh fuel. Such an operation is referred to in this work as a binary perturbation or a 'swap'. If quarter core symmetry is maintained, the perturbation translates to the other three quadrants. Likewise if eighth core symmetry is maintained, the perturbation translates to other seven octants. The question is then, how much of the core design space can binary perturbations possibly span? This work seeks to quantitatively define the space spanned by m number of binary perturbations as a function of variable size through combinatorial principles and computer simulation. The results of this study are useful in understanding the in-core fuel management problem faced by both professionals who employ hand manipulations and those who attempt stochastic solutions of the problem.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Coupled neutronic-thermal-hydraulic simulations of the European SFR core

Within the European SFR (Sodium-cooled Fast Reactor) - Safety Measures Assessment and Research Tools (ESFR-SMART) project, steady-state coupled simulation of the ESFR core has been performed using several core analysis packages, with the objective of quantifying the coupling effect. Focus is on the fuel Doppler effect and coolant expansion effect. Standalone neutronics calculations in TRACE/PARCS (PSI), DYN3D (HZDR) and WIMS (Jacobs) showed superb agreement with the reference Serpent power distribution (root mean square - rms discrepancies of 1.3%, 1.5% and 0.7% respectively). Results for COUNTHER (CIEMAT) were also in reasonable agreement, but with a somewhat higher discrepancy of 3.7%. Temperature distributions from thermal-hydraulic calculations were also compared and are found to be in good agreement. The effect of Doppler and coolant density feedback on core power distribution was predicted by TRACE/PARCS, DYN3D and WIMS to be between 0.4% and 0.8% rms difference in assembly powers. Reactivity coefficients for perturbations in the inlet temperature, flow rate and core power were shown to be negative for these three codes, with values of roughly -0.5 pcm/C. degree, -0.3 pcm/C. degree and -3.5 pcm/C. degree respectively. A preliminary investigation of differential thermal expansion effects indicates that this may have a significant effect on core power distribution of a few %, greater than anticipated a priori and may warrant inclusion in coupled core analysis to ensure the accurate calculation of power distributions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Platinum/Tantalum Carbide Core–Shell Nanoparticles with Sub‐Monolayer Shells for Methanol and Oxygen Electrocatalysis

Abstract Core–shell architectures provide great opportunities to improve catalytic activity, but achieving nanoparticle stability under electrochemical cycling remains challenging. Herein, core–shell nanoparticles comprising atomically thin Pt shells over earth‐abundant TaC cores are synthesized and used as highly durable electrocatalysts for the methanol oxidation reaction (MOR) and the oxygen reduction reaction (ORR) needed to drive direct methanol fuel cells (DMFCs). Characterization data show that a thin oxidic passivation layer protects the TaC core from undergoing dissolution in the fuel cell‐relevant potential range, enabling the use of partially covered Pt/TaC core–shell nanoparticles for MOR and ORR with high stability and enhanced catalytic performance. Specifically, at the anode the surface‐oxidized TaC further enhances MOR activity compared to conventional Pt nanoparticles. At the cathode, the Pt/TaC catalyst feature increases tolerance to methanol crossover. These results show unique synergistic advantages of the core–shell particles and open opportunities to tailor catalytic properties for electrocatalytic reactions.

Chemistry↗