Results for “Core Design” — Page 32

NASA NTRS1971

Development of a High Temperature Heater using an Yttria Stabilized Zirconia Cored Brick Matrix

The Ames pilot heater is a ceramic regenerative heater that provides high temperature air for aerodynamic and combustion experiments. The development of this heater to provide a heat storage bed with temperature capability of about 4600 R is described. A bed was designed and installed having cored brick elements of yttria-stabilized zirconia. The bed dimensions were 14 inches in diameter by 10 feet high. The thermal stress limitations of the bed were studied and maximum air flow rates based upon these limits were established. A combustion reheat system was designed and installed to provide the necessary control over the bed temperature distribution. The revised heater system was successfully operated at a maximum bed temperature of 4600 R. The successful operation demonstrated that yttria-stabilized zirconia cored brick can satisfy the high temperature-long duration requirement for storage heater applications.

Smith, K. W.↗

DOE OSTI2021

Distributed sensing of electromagnetic components using optical fiber based methods

An OFDR based fiber-optics sensor for distributed real-time temperature rise monitoring of a transformer in operation has been disclosed. The fiber-optic sensor provides an effective solution to monitoring the physical structures of the transformer core, as well as accurately detecting the non-uniform temperature distribution inside the transformer, and thus provides innovative feedback to the transformer design by minimizing the core losses. Additionally, the method may be responsive to the presence of magnetic and electric fields, as well as responsive to various chemical species. The method allows novel approaches to real-time asset monitoring of power transformers while operational.

Ohodnicki, Paul R.↗

DOE OSTI2019

Pronghorn

Pronghorn is designed as a multi-dimensional coarse-mesh conjugate heat transfer simulator (tightly coupled fluid and solid phase heat transfer) for nuclear reactor core applications. Pronghorn’s design philosophy of is that: 1. The tool will be agnostic of reactor concept, including VHTRs, SFRs, FHRs, and LWRs. 2. Must be higher fidelity than traditional nuclear systems/safety analysis (0D-1D). 3. Avoid lower-length scale phenomena, such as boundary layer theory, i.e., rely on closure relations. 4. Apply mixture theory (homogenization) wherever possible. 5. Conserve mass and energy (solve all physics on single mesh). 6. Must be easily coupled to both MOOSE-based and external physics applications. 7. Does not require a super computer. Thus, Pronghorn’s algorithmic philosophy is to resolve large scale flow features and model lower-length scale phenomena, such as boundary layers. The coarse mesh approach for lower-length scale modeling is to develop a multi-component multi-phase homogenization of solid and liquid phases into a mixture formulation inside a finite element. For example, a pebble bed core is represented as a two-phase mixture of coolant and fuel pebbles, where the fuel phase is stationary. Thus, the solid phase exerts a drag force on the fluid phase, composed of form and viscous drag, and both phases exchange internal energy through conduction and convective heat transfer. For Pronghorn, the volume fraction of fluid can vary in space, but not in time. Pronghorn has been validated against the VHTR PBMR400 benchmark problems, used for rod ejection simulations have been conducted for thermal-fluids/neutronics for both pebble-bed and prismatic gas-cooled reactors and simple thermal-neutronics coupled LWR benchmark problems. Pronghorn flow models cover a wide range of transient and steady flow conditions, including, porous fluid flow, compressible flow, and nearly incompressible, thermally expansive flow. The physics can be solved in three-dimensional Cartesian coordinates and two-dimensional cylindrical (r-z) coordinates.

Martineau, RichardC.↗

DOE OSTI2020

Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Consist v0.1.0

A Python library for provenance tracking, intelligent caching, and data virtualization in scientific simulation workflows. It automatically records code, configuration, and input data to skip redundant computations and enables querying results across many runs without manual bookkeeping. Designed to support multi-model simulation workflows like the BEAM CORE toolset at LBL, but designed to be extensible to a wide range of research workflows. Combines lineage tracking features as provided by OpenLineage with deterministic hashing like SnakeMake, and adds powerful analysis tools on model outputs.

Needell, Zachary [Lawrence Berkeley National Labor↗

DOE OSTI2025

Beyond the four core effects: revisiting thermoelectrics with a high-entropy design

Low-exergy waste heat, which constitutes the majority of industrial-scale thermal losses, remains largely unrecoverable with conventional technologies. Thermoelectrics offer a solid-state solution for converting this hard-to-access energy into electricity, making them attractive for decentralized power generation and sensor applications. High-entropy materials (HEMs) have gained traction as a strategy for better-performing thermoelectrics, but the mechanisms driving their benefits require further exploration. This article highlights key insights for heat and electronic transport in HEMs. For heat transport, we argue that reduced, and often ultralow, lattice thermal conductivity in HEMs—with respect to ordered counterparts—can be taken for granted, emerging naturally as a fifth core effect of high-entropy systems. While band convergence is often considered beneficial for electronic transport, its impact depends strongly on the electronic structure. We summarize the scenarios where it can be detrimental to thermoelectric performance. These insights motivate strategies that align seamlessly with advancements in artificial intelligence and data-driven approaches, helping accelerate the discovery of next-generation thermoelectric materials.

Oses, Corey [Johns Hopkins Univ., Baltimore, MD (U↗

NASA NTRS1973

Isotope heat source simulator for testing of space power systems

A reliable isotope heat source simulator was designed for use in a Brayton power system. This simulator is composed of an electrically heated tungsten wire which is wound around a boron nitride core and enclosed in a graphite jacket. Simulator testing was performed at the expected operating temperature of the Brayton power system. Endurance testing for 5012 hours was followed by cycling the simulator temperature. The integrity of this simulator was maintained throughout testing. Alumina beads served as a diffusion barrier to prevent interaction between the tungsten heater and boron nitride core. The simulator was designed to maintain a surface temperature of 1311 to 1366 K (1900 to 2000 F) with a power input of approximately 400 watts. The design concept and the materials used in the simulator make possible man different geometries. This flexibility increases its potential use.

Prok, G. M.↗

NASA NTRS2005

A VHDL Core for Intrinsic Evolution of Discrete Time Filters with Signal Feedback

The design of an Evolvable Machine VHDL Core is presented, representing a discrete-time processing structure capable of supporting control system applications. This VHDL Core is implemented in an FPGA and is interfaced with an evolutionary algorithm implemented in firmware on a Digital Signal Processor (DSP) to create an evolvable system platform. The salient features of this architecture are presented. The capability to implement IIR filter structures is presented along with the results of the intrinsic evolution of a filter. The robustness of the evolved filter design is tested and its unique characteristics are described.

Gwaltney, David A.↗

NASA NTRS2008

Replication of Space-Shuttle Computers in FPGAs and ASICs

A document discusses the replication of the functionality of the onboard space-shuttle general-purpose computers (GPCs) in field-programmable gate arrays (FPGAs) and application-specific integrated circuits (ASICs). The purpose of the replication effort is to enable utilization of proven space-shuttle flight software and software-development facilities to the extent possible during development of software for flight computers for a new generation of launch vehicles derived from the space shuttles. The replication involves specifying the instruction set of the central processing unit and the input/output processor (IOP) of the space-shuttle GPC in a hardware description language (HDL). The HDL is synthesized to form a "core" processor in an FPGA or, less preferably, in an ASIC. The core processor can be used to create a flight-control card to be inserted into a new avionics computer. The IOP of the GPC as implemented in the core processor could be designed to support data-bus protocols other than that of a multiplexer interface adapter (MIA) used in the space shuttle. Hence, a computer containing the core processor could be tailored to communicate via the space-shuttle GPC bus and/or one or more other buses.

Ferguson, Roscoe C.↗

DOE OSTI2021

Core-shell and egg-shell zeolite catalysts for enhanced hydrocarbon processing

Developing structure-performance relationships with the underlying goal of optimizing known zeolite catalysts involves the manipulation of their physicochemical properties. Here, we systematically assessed the impact of mesoscopic gradients in acid site concentration, which has generally received little attention in the design of zeolite catalysts for hydrocarbon upgrading. A series of core–shell MEL-type zeolites were synthesized with catalytically active ZSM-11 cores and passivated silicalite-2 shells of varying thickness. Our findings revealed that ZSM-11@silicalite-2 particles with ultrathin shells (<10 nm) have enhanced mass transport, characteristic of relatively smaller particles, compared to the corresponding ZSM-11 core. Additionally, catalytic testing using the methanol-to-hydrocarbon (MTH) reaction showed that core–shell zeolites exhibit longer lifetimes, higher total turnovers, and an unexpected promotion of the aromatic cycle in the hydrocarbon pool mechanism. Time-resolved acid titration of core and core–shell catalysts confirmed that the siliceous shell introduces a hydrophobic exterior that impacts molecular diffusion. In comparison, prepared MFI core-shells (ZSM-5@silicalite-1) showed similar enhancement in catalyst performance. Moreover, we prepared egg-shell configurations of each zeolite, silicalite-2@ZSM-11 and silicalite-1@ZSM-5, comprised of an inert core and catalytically active shell. This inverse design of the egg-shell created pseudo nanosheets with total turnovers that were markedly higher than their homogeneous counterparts. Collectively, this study demonstrated that mesoscopic gradients in acid concentration via the design of core–shell and egg-shell zeolites significantly improve catalyst performance over conventional analogues for hydrocarbon upgrading.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2025

Griffin Capability Improvements in Support of Ex-core Deep-Penetration Problems

Advanced reactor designs, especially portable reactors that are designed to be located closer to humans and operate autonomously, require the ability to accurately compute the ex-core neutron and gamma flux solutions in terms of shielding design optimization to reduce dose rates at the vessel boundary and detector signal prediction to drive the reactor control system. The Nuclear Energy Advanced Modeling and Simulation program has prioritized improvements to the Griffin discrete ordinates (SN) solver for deep-penetration problems in fiscal year 2025. Significant advancements have been made to the Griffin methodologies for solving ex-core deep-penetration problems for steady-state, fixed-source and transient calculations. This work presents the methodology improvements as well as a comprehensive demonstration with a Transient Test Reactor model and measurements.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Solid-state laser refrigeration of core-shell polystyrene microspheres

Microlaser designs based on the coupling of whispering gallery modes (WGMs) with the upconversion processes which take place within lanthanide-doped nanoparticles (UCNPs) have been demonstrated and shown to have many valuable qualities, such as high Q factors and low lasing thresholds. One obstacle that these microlaser designs still face is the challenges caused by photothermal heating of the gain medium, which could be solved through the design of a radiation balanced microlaser. In this work, WGM microresonators composed of 5 µm diameter polystyrene spheres are fabricated with a layer of Yb3+-doped NaYF4 UCNPs in order to test if the anti-Stokes cooling properties of the UCNPs can cool the microresonator and its environment under laser irradiation. We find via calibrated mean fluorescence spectroscopy that the UCNPs can cool their local environment by as much as 23 °C and significantly reduce the heating of the aqueous environment surrounding the microresonator, showing promise for inclusion in a design for a radiation balanced microlaser.

whispering gallery modes, optical refrigeration, m↗

DOE OSTI2019

Neutron Absorber Considerations for the DOE Standardized Canister

The Department of Energy (DOE) standardized canister is being re-evaluated to support the management of non-commercial spent nuclear fuel. The standardized canister would allow a broad range of DOE-managed fuels to be loaded, stored, transported, and disposed of while avoiding the need to re-open the canister after initial loading, thus preventing repackaging of the material. A key safety function of this package is to ensure the material maintains a sub-critical configuration for storage, transportation, and disposal. DOE-managed fuels come from a wide range of reactor types, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. To adequately address the diversity of DOE spent nuclear fuel (SNF), a number of basket were designed. The basket material needed to have the ability to ensure subcriticality, be structurally sound, and be corrosion resistant. One such option that was investigated in the late-1990s through mid-2000s was a corrosion-resistant, nickel-chromium-molybdenum alloy containing gadolinium, which was termed the advanced neutron absorber (ANA). A number of analyses were performed to verify the structural, corrosive properties, and neutron absorption performances of this material as well as to compare other potential materials to be used for the basket. Gadolinium-containing carbon steel and borated stainless steel were also considered and evaluated. Boral was briefly considered, but due to its corrosive properties. Each of the evaluated materials have strengths and weaknesses. ANA is the best thermal neutron absorber but lacks confirmatory research for corrosion and weldability. Gadolinium containing carbon steel provides excellent moderator exclusion capabilities but does not contain a sufficient supply of phosphate to ensure the gadolinium precipitates. To minimize the probability of a critical configuration, it is highly desirable for the gadolinium to precipate the probability that the materials will remain within the waste packages over the postclosure time periods of interest. Borated stainless steel has structural integrity and impressive corrosion resistance in most environments, but can undergo galvanic corrosion when in close proximity/contact with aluminum and may experience less than desirable corrosion resistance. This report provides an extensive list and comparison of the materials evaluated to ensure sub-criticality of DOE standardized canister in all credible scenarios. More analyses need to be performed using new and updated corrosion information to re-evaluate the neutron absorber that best meets the objectives of the DOE standardized canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW PPT

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types, such as light- and heavy-water-moderated reactors, graphite-moderated reactors, and breeder reactors, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. Because of the wide variety and condition of SNF, a robust canister termed the DOE Standardized SNF Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard SNF Canister had four variations consisting of a 10- or 15-foot length and an 18- or 24-inch diameter. For ultimate disposal in the Yucca Mountain Repository these canisters were to be grouped with 24-inch diameter HLW canisters in a codisposal waste package. The 18-inch DOE Standardized SNF Canister could be placed in the middle of five HLW canisters. The 24-inch DOE Standardized SNF Canister would take the place of one the five HLW canister on the outer ring in the codisposal waste package. No DOE Standardized SNF Canisters have been loaded, and many fuel types can be loaded in multiple canisters. Depending on the selected diameter of the DOE Standardized SNF Canister for a particular fuel, the number of potential codisposal waste packages could vary. In addition, different types of DOE SNF may be able to be combined in a single canister. This could significantly change the number of DOE SNF canisters that are loaded and result in a decrease of the number of total potential codisposal waste packages. Ultimately, if too many HLW canisters need disposition, then a reduction in the number of DOE Standardized SNF Canisters will change the final total of codisposal waste packages very little. Likewise, if the ratio of DOE Standardized SNF Canisters to HLW canister produced is too high, then DOE Standardized SNF Canisters will not have enough HLW canisters to surround them in the codisposal waste package. This paper examines the results of differing loading strategies for a few DOE SNFs. It also compares the ranges of DOE Standardized SNF Canisters that may be produced to the ranges of HLW canisters that may be produced in order to determine potential scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Co-Disposal Waste Package Loading Options for DOE SNF and HLW - 20331

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1976

Mechanical design

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.

Source record↗

NASA NTRS1979

Minimum-mass designs of stiffened graphite/polyimide compression panels

This study presents results from an analytic investigation to determine minimum-mass designs of stiffened Gr/Pi compression panels for a wide range of uniaxial compression loads. Four panel configurations are considered: (1) hat stiffened-laminated skin, (2) hat stiffened-honeycomb core sandwich skin, (3) blade stiffened-laminated skin and (4) blade stiffened-honeycomb core sandwich skin panels. Designs are generated by an automatic optimization computer code entitled PASCO (Panel Analysis and Sizing COde). The analysis portion of PASCO contains a generalized buckling analysis which calculates all critical modes without assuming any of them a priori. Mass-strength curves are presented for the four configurations along with specific designs generated for low, moderate and high values of compression load. A mass-strength curve for a titanium hat stiffened panel is provided for comparison. A 30% to 50% mass savings over titanium panels is shown for moderately and heavily loaded Gr/Pi panels.

Weaver, G. G., II↗