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

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At least 1,423 records · Page 79
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Collective memory transfer devices and methods for multiple-core processors

This disclosure provides systems, methods, and apparatus for collective memory transfers. A control unit may be configured to coordinate a transfer of data between a memory and processor cores. For a read data transfer operation, the control unit may receive a trigger packet identifying a read data transfer operation and identifying a first plurality of data lines based on data values included in the trigger packet. The control unit may read the first plurality of data lines from the memory sequentially and send a second plurality of data lines to the processor cores. For a write data transfer operation, the control unit may send a request for at least one data line to a plurality of processor cores, may receive and reorder the requested data lines, and may write the data lines to a memory. The control unit may determine a mapping between processor cores and the memory.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Core-shell structured nanoparticles for lithium-sulfur cells

Described is a core-shell nanoparticle comprising a lithium sulfide nanoparticle core and a shell covering the lithium sulfide nanoparticle core, the shell comprising at least one of carbon, polyanaline or a transition metal sulfide. The core-shell nanoparticle may be used for a positive electrode in a lithium/sulfur battery cell.

25 ENERGY STORAGE↗
DOE OSTI2020

Lubricant compositions comprising core-shell nanoparticles

Lubricant compositions, core-shell nanoparticles, and related methods are disclosed. In an exemplary embodiment, a lubricant composition includes a plurality of core-shell nanoparticles. The nanoparticles include a core, a first shell disposed on the core, and a second shell disposed on the first shell. The first shell is formed from a siliceous material and the second shell is formed from a hydrophobic material. The first and second shells form functional coatings that reduce wear and friction of parts lubricated with the lubricant composition.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2017

What Has Been Learned From Pressure Cores

The advancement of pressure core acquisition and analysis technology in recent decades has enabled detailed imaging and direct measurement of naturally occurring hydrate-bearing sediments and has shed light onto hydrate habits, formation processes, fundamental physical properties, and hydrate deposit responses during gas production. This paper reviews the development and capabilities of the pressure core technology, discusses insights learned from hydrate-bearing pressure cores and inherent limitations of this technology, and concludes with suggestions for potential applications of pressure core technology for investigating hydrate deposits.

Dai, Sheng↗
DOE OSTI2021

Nuclear Requalification of the New ATR Core

The Advanced Test Reactor (ATR) creates a unique high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector eventually requiring a replacement during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data to model and predict criticality scenarios for the new unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult to find data trends that could be useful for the modeling and prediction process. So nuclear requalification testing is conducted using different flux and irradiation tests to accurately verify the models and to recalibrate nuclear monitoring instrumentation in the ATR core allowing for safe re-start of normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Core-shell structured nanoparticles for lithium-sulfur cells

Described is a core-shell nanoparticle comprising a lithium sulfide nanoparticle core and a shell covering the lithium sulfide nanoparticle core. The core-shell nanoparticle may be used for a positive electrode in a lithium/sulfur battery cell.

Cairns, Elton J.↗
DOE OSTI2021

Heat pipe reactor core and heat exchangers formation and deployment

A heat pipe reactor may include a reactor core and one or more heat exchangers positioned on one or both sides of the reactor core. The heat pipe reactor may also include a plurality of heat pipes extending from the reactor core and out through the one or more heat exchangers. The reactor core may be composed of a plurality of monolithic blocks.

Reid, Robert Stowers↗
DOE OSTI2023

CMSSW Scaling Limits on Many-Core Machines

Today the LHC offline computing relies heavily on CPU resources, despite the interest in compute accelerators, such as GPUs, for the longer term future. The number of cores per CPU socket has continued to increase steadily, reaching the levels of 64 cores (128 threads) with recent AMD EPYC processors, and 128 cores on Ampere Altra Max ARM processors. Over the course of the past decade, the CMS data processing framework, CMSSW, has been transformed from a single-threaded framework into a highly concurrent one. The first multithreaded version was brought into production by the start of the LHC Run 2 in 2015. Since then, the framework's threading efficiency has gradually been improved by adding more levels of concurrency and reducing the amount of serial code paths. The latest addition was support for concurrent Runs. In this work we review the concurrency model of the CMSSW, and measure its scalability with real CMS applications, such as simulation and reconstruction, on mode rn many-core machines. We show metrics such as event processing throughput and application memory usage with and without the contribution of I/O, as I/O has been the major scaling limitation for the CMS applications.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Computational and Experimental Examination of Simulated Core Damage and Relocation Dynamics of a BWR Fuel Assembly - 16013

The Japan Atomic Energy Agency (JAEA) plans a large-scale test to evaluate damage and relocation behavior of Boiling Water Reactor (BWR) core materials consisting of fuel rods (zirconia as fuel simulant), channel boxes, control blade assemblies as well as lower core-support structures. Its purpose is to contribute to the understanding of the core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing anticipated at JAEA in the future, JAEA desires preliminary investigations to examine melting test pieces, being composed of a mixture of zirconia and metals by a non-transferred thermal plasma heating system. The purpose of such tests is to verify the materials and test piece will be heated by plasma as a primary heat source to the target temperature of ca. 2900 K and to collect data about the relocation behavior of the high-temperature materials. This paper describes the first in a series of tests demonstrating the effective use of thermal plasma technology to recreate the dynamics of a typical fuel core bundle channel assembly section. Results from preliminary computational simulations of simplified fuel bundle assemblies are presented illustrating the effectiveness of a 150 kW, non transferred (NTR) configuration of a thermal plasma argon jet to heat the assembly within a highly-insulated containment vessel. An experimental test program using the computational analyses as a basis and a Phoenix Solutions Co (PSC) commercial model PT150, 150 kW plasma torch is described and the test results are presented illustrating the time versus temperature history of the fuel bundle assembly melting dynamics. Details on plasma operation are discussed including: power requirements, current flow, voltage levels achieved, argon gas flow rates, processing times and standoff distances of the torch discharge from the top of the fuel bundle assembly (achieved with remote control of the axial (vertical) translation of the torch within the containment processing chamber). Photos taken from the top of the processing chamber are presented illustrating the various phases of melting characteristics of the zircaloy-clad, zirconium oxide filled fuel tubes, all within a stainless steel tube bundle channel centered within an insulated zirconia crucible. Based on lessons learned from this initial test campaign, a brief discussion of the requirements and features of a follow-on Phase II test program is provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Assessment of Near-Breeding Heterogeneous Seed/Blanket Cores in Pressure-Tube Heavy Water Reactors with Thorium-Based Fuels

To help extend uranium resources, thorium, which is three to four times as abundant as uranium, can be exploited as an alternative fertile nuclear fuel. The pressure tube heavy water reactors (PT-HWR) is an excellent mainstream technology for harnessing the energy potential in thorium due to its high neutron economy enabled by using heavy water as a moderator and coolant, and by refueling online. The PT-HWR is also capable of flexibility in fuel management, allowing the use of multiple fuel types with different refueling rates. One approach to using thorium-based fuels in PT-HWRs is to use a homogeneous core with a once-through thorium (OTT) fuel cycle, where a fissile material is mixed with thorium (such as (LEU,Th)O{sub 2} or (Pu,Th)O{sub 2}), burned in a high-conversion reactor, and then put into storage for several decades until it is economically attractive to reprocess and recycle. An alternative approach is to use a heterogeneous seed/blanket core, where the seed/driver fuel is made of ({sup 233}U,Th)O{sub 2}, and the blanket fuel is make of ThO{sub 2}, and attempt to achieve a very-high conversion ratio (CR) core, and perhaps net breeding with a self-sufficient equilibrium thorium (SSET) fuel cycle. The objective of this study was to assess the performance of various heterogeneous seed/blanket cores in a 700-MWe-class PT-HWR using ({sup 233}U,Th)O{sub 2} fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Microstructure of In-Core Molten Salt Corrosion Hastelloy N{sup R} and 316 Stainless Steel

Since 2011, many materials have been discussed for potential application in Fluoride salt-cooled High-temperature nuclear Reactors (FHRs), which are emerging as leading reactor concept for the next generation of nuclear reactors thanks to the benefits of a high degree of passive safety, high thermal efficiency, and low spent fuel volume. Among these discussed materials, Hastelloy N{sup R} and 316L stainless steel have been selected as the most likely candidate alloys for FHRs. To provide direct experience in integrating a real system using molten FLiBe salt for reactor operation, these two candidate alloys were experimentally tested in Li-7 enriched Li{sub 2}BeF{sub 4} (FLiBe) molten salt in the core of MIT Research Reactor (MITR) at 700 deg. C for 1000 hours. After in-core corrosion testing, some basic post-irradiation examinations (PIE) including weight change per unit area and surface profile as well as surface morphology were conducted in the MIT Nuclear Reactor Laboratory (NRL) and compared with out-of-reactor corrosion samples. The deep understanding of these alloys in high-temperature molten salt with strong neutron irradiation requires further microstructural analysis such as electron microscopy and energy dispersive spectroscopy on the surface and cross-section of the in-core corrosion tested samples. The challenge in the microstructural analysis is due to the high radioactivity from the gamma emitting isotopes formed in alloys during neutron irradiation. Additionally, the facilities that are authorized for radioactive materials are very limited on campus. In this study, the microstructure of the in-core corrosion tested Hastelloy N{sup R} and 316L stainless steel specimen was analyzed by using the facilities at the Center for Advanced Energy Studies (CAES) at Idaho National Laboratory (INL) through the Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Deterministic Selection of Design Extension Condition without Core Melt

The concept of beyond design basis accidents (BDBA) recently has been changed to design extension conditions (DEC). DEC is required to be considered in the design process, and it comprises conditions in events without core melt and conditions in events with core melt. DEC with core melt is a part of severe accident. A systematic and deterministic approach to select DEC without core melt for a pressurized water reactor (PWR) is introduced.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Comparisons of Nodal diffusion and whole-core transport methods for multiple cycles of a small Light water reacto

As part of a cooperative research and development agreement between the US department of Energy's Oak Ridge National Laboratory and Holtec International subsidiary SMR LLC, core simulator models of the SMR-160, a light water-cooled and -moderated, natural circulation, small modular reactor, were developed and compared. The model results indicate pin-by-pin power histories, critical boron concentration, and other key performance indicators useful for developing multi-cycle core loading patterns. One model uses industry standard nodal methods with a simple thermal-hydraulics solution, and the other uses an advanced deterministic neutronics solver with a coupled subchannel calculation. The goals of the comparison are to provide the reactor designer with a detailed and coupled core physics reference solution, inform the reactor designer of nodal model best practices based on the code-to-code comparison, and set up a workflow for performing subsequent detailed core simulator calculations. The results of this work will be used in the development of coupled neutronic, thermal hydraulic, and fuel performance models for advanced analysis of the SMR-160. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Nuclear Requalification of the ATR Core

The Advanced Test Reactor (ATR) creates a unique, high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector. The reflector is eventually required to be replaced during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data, notably 1977-1994 CIC data, to conduct predictive criticality analysis on the new, unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult, and in some cases impossible, to utilize. These inconsistencies arise in various forms of not using correct units, inconsistent diagram usage, and not portraying the same kind of data similarly across the reports. To work around this inconsistency issue, Reactor Engineering is implementing the use of several new neutronic modeling codes that utilize modern computation speeds and methods to model the new core in different criticality environments. Further validation is then conducted by nuclear testing and bringing the reactor critical and conducting various irradiation tests. Upon completion and certification that the reactor passed all testing requirements, ATR can once again re-start normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Validation of HELIOS for ATR Core Follow Analyses

This work summarizes the validation analyses for the HELIOS code to support core design and safety assurance calculations of the Advanced Test Reactor (ATR). Past and current core safety assurance is performed by the PDQ-7 diffusion code; a state of the art reactor physics simulation tool from the nuclear industry’s earlier days. Over the past twenty years, improvements in computational speed have enabled the use of modern neutron transport methodologies to replace the role of diffusion theory for simulation of complex systems, such as the ATR. More exact methodologies have enabled a paradigm-shift away from highly tuned codes that force compliance with a bounding safety envelope, and towards codes regularly validated against routine measurements. To validate HELIOS, the 16 ATR operational cycles from late-2009 to present were modeled. The computed power distribution was compared against data collected by the ATR’s on-line power surveillance system. It was found that the ATR’s lobe-powers could be determined with ±10% accuracy. Also, the ATR’s cold startup shim configuration for each of these 16 cycles was estimated and compared against the reported critical position from the reactor log-book. HELIOS successfully predicted criticality within the tolerance set by the ATR startup procedure for 13 out of the 16 cycles. This is compared to 12 times for PDQ (without empirical adjustment). These findings, as well as other insights discussed in this report, suggest that HELIOS is highly suited for replacing PDQ for core safety assurance of the ATR. Furthermore, a modern verification and validation framework has been established that allows reactor and fuel performance data to be computed with a known degree of accuracy and stated uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Utah FORGE: Deep Well 58-32 (MU-ESW1) Core Data

These datasets, images, and graphics were derived from core drilling and core that was extracted from Utah Forge deep well 58-32 (originally called MU-ESW1), near Roosevelt Hot Springs.

15 GEOTHERMAL ENERGY↗
DOE OSTI2014

Simultaneous in-situ synthesis and characterization of Co@Cu core-shell nanoparticle arrays

Core-shell nanostructures have attracted much attention due to their unique and tunable properties relative to bulk structures of the same materials, making core-shell nanoparticles candidates for a variety of applications with multiple functionalities. [1,2] Intriguing magnetic behavior can be tailored by variation of size, interface, crystal orientation, and composition, and core-shell nanostructures with noble-metal shells yield novel optical responses [3] and enhanced electrocatalytic activity. [4]

77 NANOSCIENCE AND NANOTECHNOLOGY↗