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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
NASA NTRS1996

Induced radioactivity of LDEF materials and structural components

We present an overview of the Long Duration Exposure Facility (LDEF) induced activation measurements. The LDEF, which was gravity-gradient stabilized, was exposed to the low Earth orbit (LEO) radiation environment over a 5.8 year period. Retrieved activation samples and structural components from the spacecraft were analyzed with low and ultra-low background HPGe gamma spectrometry at several national facilities. This allowed a very sensitive measurement of long-lived radionuclides produced by proton- and neutron-induced reactions in the time-dependent, non-isotropic LEO environment. A summary of major findings from this study is given that consists of directionally dependent activation, depth profiles, thermal neutron activation, and surface beryllium-7 deposition from the upper atmosphere. We also describe a database of these measurements that has been prepared for use in testing radiation environmental models and spacecraft design.

NASA Discipline Radiation Health↗
DOE OSTI2019

Challenges of SLOWPOKE-2 HEU Irradiated Fuel Transportation from Canada to USA - 19123(2)

The SLOWPOKE-2 (Safe Low-Power Kritical Experiment) was designed by AECL (Atomic Energy of Canada Limited) in the late 1960's. The core consists of HEU (highly enriched uranium) fuel with an enrichment of 93%. The fuel cage typically has approximately 300 fuel pins assembled in a triangular pitch between the two plates. The fuel pins are made from uranium-aluminum alloy cladded with aluminum. The SLOWPOKE-2 reactor is rated at a nominal 20 kW. with varying neutron flux, maximum being 1 E12 n/cm{sup 2}/s. The mass of the fuel cage is 5.06 kg. The life of the core can be extended by adding layers of beryllium reflector. The transportation flask for the irradiated fuel is a Type B(U)F. The shipment classification is UN3328 Class 7. To date, 4 SLOWPOKE-2 reactors have been decommissioned. The fuel from all these reactors was transported to the U.S. The CNSC (Canadian Nuclear Safety Commission) joint Canada-United States guide for approval of Type B(U) and fissile material transportation packages (RD-364) satisfies the requirements stipulated in International Atomic energy Agency (IAEA) TS-R-1 1996 edition (revised) and is acceptable to the DOT (Department of Transportation) and NRC (Nuclear Regulatory Commission) for complying with U.S. code of federal regulations 10 CFR Part 71. The export license is issued by the CNSC. The transportation licence is issued by CNSC after approval from TC (Transport Canada) and also by DOT after approval from NRC. The final home of the HEU SLOWPOKE-2 irradiated fuel is in Savannah River Site of DOE (Department of Energy). The transportation licence requires that the Type B(U)F flask is structurally sound in case of a crash. The US DOT provides a CAC (Competent Authority Certificate) which describes the irradiated fuel specifications that can be transported in the designated flask. The U.S. DOT also requires a leak test certification in accordance with ANSI N14.5-2014 to verify a leakage rate not to exceed 1 E-7 std-cm{sup 3}/s. As most SLOWPOKE-2 facilities are not located on the existing approved routes for nuclear fuel transportation, a route survey needs to be performed and submitted to the NRC. The route survey report is assessed by the NRC and issues a route approval for the transportation of the fissile material. This technical paper will discuss the detailed processes required to obtain approvals in order to transport the HEU SLOWPOKE-2 irradiated fuel from Canada to U.S. based upon recent experiences of decommissioning SLOWPOKE-2 reactors in Canada. It will provide details on the requirements of various regulators in Canada and the U.S. that govern and regulate this shipment including transportation, safety and security. It will also present lessons learned from previous experiences that can be utilized for best practice on other similar projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Fluor Idaho Technical Investigation-ARP V Lid Ejection Reactants and Mechanism - 19521

In April 2018, four 208.2-L (55-gal) drums containing transuranic waste at the Idaho National Laboratory Site became over-pressurized, resulting in lid ejection and expulsion approximately 50% of the contents. The parent drums were part of a retrieval effort and were exhumed from a portion of Subsurface Disposal Area Pits 11 and 12 beginning in 1974 and ending in 1978. Drums were later retrieved from the cargo containers, characterized using real-time radiography and nondestructive assay, and opened. The contents were visually inspected and repackaged into new 'daughter' drums within the Accelerated Retrieval Project V facility. Lid ejection occurred 7-12 hours after repackaging. The lid ejections were accompanied by significant heat generation. Empirical evidence of heat generation was observed in each reacted drum. The heat generated was sufficient to warp the high-density polyethylene liner (< 122 to 137 deg. C) in three of the drums and char the drum paint (∼400 deg. C) on one drum. To determine the potential reactant(s), a comprehensive chemical analysis was performed by two independent laboratories, Southwest Research Institute and Savannah River National Laboratory, on the reacted drum contents, ejected drum material, and related drums. A comprehensive chemical compatibility analysis was performed, and the most likely reactant(s) were identified. The oxidation of depleted uranium was the source of the heat generation. The presence of depleted uranium in the daughter drums and its chemical behavior (reactive properties and lack of visual indicators during packaging) are consistent with the event. The heat generated was sufficient to accelerate secondary chemical reactions leading to significant gas formation sufficient to generate pressures well in excess of the pressures needed to eject the lid and contents. These reactions fall into two categories: (1) simple volatilization of organics and water present in the drum and (2) hydrolysis of beryllium carbide, which generated significant quantities of methane gas. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Early glacier advance in New Zealand during the Antarctic Cold Reversal

ABSTRACT Glacial landscapes preserve records of past climate change. Investigating the glacier–climate system over the Late Quaternary provides information about past climate change and context for present‐day glacier response to climate warming. Using 28 beryllium‐10 ( 10 Be) surface exposure dates and snowline reconstructions, we present glacier fluctuations and climate changes for the Antarctic Cold Reversal in the Ahuriri River catchment, Southern Alps of New Zealand (44°7′50″S, 169°38′29″E). Prominent terminal and lateral moraine features from the upper right tributary of the Ahuriri River valley have exposure ages of 14.5 ± 0.3, 13.6 ± 0.3 and 12.6 ± 0.2 ka, suggesting retreat of the glacier during the Antarctic Cold Reversal. Maximum elevation of lateral moraines (MELM) and accumulation area ratio (AAR) suggest snowline elevations at these ages were ≤700, ≤630 and ~360 m lower than today, respectively. This equates to air temperatures ≤3.9, ≤3.5 and 2.3 ± 0.7 °C lower than today (1981–2010), assuming no changes in past precipitation. Ice‐sculpted bedrock surfaces bound by a lateral moraine at nearby Canyon Creek have an age of 13.1 ± 0.3 ka, indicating the moraine correlates with those in the Ahuriri upper right tributary. MELM and AAR reconstructions from the Canyon Creek suggest that snowline elevations at 14.5–13.6 ka were ≤500 or ~380 m lower than today, corresponding to air temperatures ≤2.8 or 2.4 ± 0.7 °C lower than the present‐day (1981–2010). Our results provide insight into the structure of the Antarctic Cold Reversal in the Southern Alps, showing that the largest glacier advance occurred at the start of this interval at c . 14.5 ± 0.3 ka and was followed by gradual retreat. We hypothesize that the early cooling and glacier readvance in New Zealand at the onset of the Antarctic Cold Reversal were triggered by a latitudinal shift of the Southern Hemisphere westerly wind belt.

58 GEOSCIENCES↗
DOE OSTI2024

Bayesian batch optimization for molybdenum versus tungsten inertial confinement fusion double shell target design

Access to reliable, clean energy sources is a major concern for national security. Much research is focused on the “grand challenge” of producing energy via controlled fusion reactions in a laboratory setting. For fusion experiments, specifically inertial confinement fusion (ICF), to produce sufficient energy, the fusion reactions in the ICF fuel need to become self-sustaining and burn deuterium-tritium (DT) fuel efficiently. The recent record-breaking NIF ignition shot was able to achieve this goal as well as produce more energy than used to drive the experiment. This achievement brings self-sustaining fusion-based power systems closer than ever before, capable of providing humans with access to secure, renewable energy. In order to further progress toward the actualization of such power systems, more ICF experiments need to be conducted at large laser facilities such as the United States's National Ignition Facility (NIF) or France's Laser Mega-Joule. The high cost per shot and limited number of shots that are possible per year make it prohibitive to perform large numbers of experiments. As such, experimental design relies heavily on complex predictive physics simulations for high-fidelity “preshot” analysis. These multidimensional, multi-physics, high-fidelity simulations have to account for a variety of input parameters as well as modeling the extreme conditions (pressures and densities) present at ignition. Such simulations (especially in 3D) can become computationally prohibitive to turn around for each ICF experiment. In this work, we explore using Bayesian optimization with Gaussian processes (GPs) to find optimal designs for ICF double shell targets, while keeping computational costs to manageable levels. These double shell targets have an inner shell that grades from beryllium on the outer surface to the higher Z material molybdenum, as opposed to the nominally used tungsten, on the inside in order to trade off between the high performance associated with high density inner shells and capsule stability. We describe our results for “capsule-only” xRAGE simulations to study the physics between different capsule designs, inner shell materials, and potential for future experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

High Flux Isotope Reactor (HFIR)

Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) is the highest-flux reactor-based neutron source in the United States. HFIR’s intense neutron flux and state-of-the-art facilities result in world-class capabilities, including neutron scattering, radioisotope production, materials irradiation, and neutron activation analysis. HFIR’s multimission capabilities are attributed to its versatile high-power-density core design consisting of a series of concentric regions, including a flux trap target region, an inner fuel element, an outer fuel element, a control element region, and a beryllium reflector. The pressurized, light-water-cooled research reactor operates at 85 MW and is fueled by 9.4 kg 235U.

Chandler, David↗
DOE OSTI2022

Be Surface Structures on W(110) and W(211): A DFT Study

Beryllium and Tungsten are promising candidates for use as plasma facing materials (PFMs) in upcoming fusion reactors. Many complex and competing phenomena however complicate the understanding and development of materials that operate in the harsh conditions of the reactor. In particular, redeposition of Be onto the W divertor must be considered due to the expected erosion of the Be first wall under exposure to the plasma. It is known that a build up of Be on W allows for the formation of BeW alloys which can harm the longevity and performance of the fusion divertor. In an effort to understand the interaction of Be with W surfaces, a study of Be structures on W(110) and W(211) as a function of Be coverage has been carried out using Density Functional Theory. We have found that both W surfaces develop monolayers of Be characterized by a densely packed hexagonal structure. Below this monolayer coverage, Be structures are found to have two motifs of coverage, a densely packed hexagonal pattern punctuated by areas of low-density coverage. The structures found here produce work function values and trends in good agreement with experimental measurements.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Modeling a generic TRISO-fueled heat pipe microreactor using SCALE: Depletion, transportation criticality, and shielding

This paper demonstrates the applicability of the SCALE code system to tristructural-isotropic (TRISO)-fueled heat pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study conducted supports US Nuclear Regulatory Commission code readiness efforts for advanced non–light-water reactor technologies and is intended as a code capability demonstration rather than as an optimization of a specific microreactor design. The modeled reactor employs high-assay low-enriched uranium (HALEU) uranium oxycarbide (UCO) TRISO fuel and beryllium oxide (BeO) reflectors and operates at 7.5 MWth with a nominal lifetime of about 3 effective full power years. Representative cases for fresh and irradiated cores were selected to exercise SCALE methods relevant to reactor operation and post-irradiation transport. The discharged-core decay heat is approximately 6% of operating power immediately after shutdown. Transportation criticality calculations show that internal water ingress is the dominant reactivity effect, with fully flooded fresh core and irradiated core configurations remain above the subcriticality criterion, even with the available control mechanisms. Shielding calculations for a simplified transportation package indicate that normal-condition dose rates are governed mainly by shielding thickness and cooling time, whereas the breached hypothetical accident case is governed primarily by cooling time. Overall, the study shows that SCALE supports depletion, transportation criticality, and shielding evaluations efficiently for TRISO-fueled heat pipe microreactors within a single code system.

Criticality↗
DOE OSTI2022

Application of the Redlich-Kister expansion for estimating the density of molten fluoride psuedo-ternary salt systems of nuclear industry interest

The development of next-generation molten salt reactors relies on accurate knowledge of the thermophysical properties of the candidate coolant and fueled molten salts. These thermophysical properties include density, viscosity, thermal conductivity, and heat capacity. Because of difficulties in measuring thermophysical properties of molten salts, there are many gaps in the current state of thermophysical property knowledge of these salts, particularly those that contain actinides or beryllium. Therefore, leveraging modeling techniques to estimate unknown molten salt thermophysical properties and guide future experimental measurements has high value for the nuclear industry. Here, the densities of molten fluoride pseudo-ternary salt systems, which are of interest to the nuclear industry, were estimated using Redlich-Kister expansion and Muggianu interpolation techniques. The pseudo-ternary systems considered for estimation in this study were NaF-LiF-ZrF 4 , LiF-BeF 2 -ZrF 4 , LiF-BeF 2 -ThF 4 , NaF-LiF-BeF 2 , NaF-KF-BeF 2 , NaF-ZrF 4 -UF 4 , and NaF-BeF 2 -UF 4 . This Redlich-Kister estimation approach accounts for nonideal mixing behavior based on pseudo-binary subsystem interaction parameters determined from experimentally measured pseudo-binary system density data sets. The Redlich-Kister estimation was compared with the method of additive molar volumes, which assumes ideal mixing. Additionally, the Redlich-Kister approach was used to determine previously unknown binary and ternary interaction parameters based on experimentally measured density data sets for select pseudo-ternary salt systems. The results of this study show improvement in density estimation using the Redlich-Kister approach for all systems considered compared with estimation by additive molar volumes. Furthermore, this analysis allowed for the estimation of nonideal density behavior in experimentally unstudied ZrF 4 -UF 4 and BeF 2 -UF 4 , as well as the quantification of ternary interaction in NaF-LiF-ZrF 4 , NaF-BeF 2 -UF 4 , and NaF-ZrF 4 -UF 4 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

A quantum Monte Carlo study of systems with effective core potentials and node nonlinearities

In this report we study beryllium dihydride (BeH2) and acetylene (C2H2) molecules using real-space diffusion Monte Carlo (DMC) method. The molecules serve as perhaps the simplest prototypes that illustrate the difficulties with biases in the fixed-node DMC calculations that might appear with the use of effective core potentials (ECPs) or other nonlocal operators. This is especially relevant for the recently introduced correlation consistent ECPs (ccECPs) for 2s2p elements. Corresponding ccECPs exhibit deeper potential functions due to higher fidelity to all-electron counterparts, which could lead to larger local energy fluctuations. We point out that the difficulties stem from issues that are straightforward to address by upgrades of basis sets, use of T-moves for nonlocal terms, inclusion of a few configurations into the trial function and similar. The resulting accuracy corresponds to the ccECP target (chemical accuracy) and it is in consistent agreement with independent correlated calculations. Further possibilities for upgrading the reliability of the DMC algorithm and considerations for better adapted and more robust Jastrow factors are discussed as well.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Materials Learning Algorithms (MALA): Scalable machine learning for electronic structure calculations in large-scale atomistic simulations

We present the Materials Learning Algorithms (MALA) package, a scalable machine learning framework designed to accelerate density functional theory (DFT) calculations suitable for large-scale atomistic simulations. Using local descriptors of the atomic environment, MALA models efficiently predict key electronic observables, including local density of states, electronic density, density of states, and total energy. The package integrates data sampling, model training and scalable inference into a unified library, while ensuring compatibility with standard DFT and molecular dynamics codes. We demonstrate MALA's capabilities with examples including boron clusters, aluminum across its solid-liquid phase boundary, and predicting the electronic structure of a stacking fault in a large beryllium slab. Scaling analyses reveal MALA's computational efficiency and identify bottlenecks for future optimization. With its ability to model electronic structures at scales far beyond standard DFT, MALA is well suited for modeling complex material systems, making it a versatile tool for advanced materials research.

Density functional theory↗