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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Emulating Variable Uranium Enrichments With Radiation Signature Training Devices (RSTD)

Radiological signature training devices (RSTDs) are sealed sources (or assemblies of sources) that emulate the radiation signature of larger masses of special nuclear material for passive detection while maintaining a minimal logistical footprint in security, radiological safety, criticality safety, and transportation requirements. Previous work focused on emulation of large quantities of highly enriched uranium for search and field identification with radioisotope identification detectors and mobile systems. New work has expanded into investigating use of the RSTDs with laboratory safeguards tools, such as MGAU and FRAM, and emulating uranium spectra of multiple enrichments over more energy regions (particularly the x-ray region).

Blessinger, Christopher S.↗
DOE OSTI2022

Methods of producing enriched scandium-47, and related systems and apparatuses

A method of producing enriched 47 Sc comprises irradiating a V structure comprising 51 V with at least one incident photon beam having an endpoint energy within a range of from about 14 MeV to about 44 MeV to convert at least some of the 51 V to 47 Sc and form a 47 Sc-containing structure. The 47 Sc of the 47 Sc-containing structure is separated from additional components of the 47 Sc-containing structure using a chromatography process. Systems and apparatuses for producing enriched 47 Sc are also described.

Snow, Mathew S.↗
DOE OSTI2021

Enrichments of Anaerobic Selenium Oxyanion Reducers from FGD Wastewater Effluent

Flue gas desulfurization systems (FGD) mitigate air pollutants from coal fired power plants. However, such systems produce wastewater that contains other pollutants such as mercury, arsenic, nitrogen, and selenium that must be removed before the water is released back into the environment. Selenium is of interest because chronic low doses of selenium (ng/L) can cause detrimental ecological consequences. Biological treatment can be utilized to remove selenium from FGD wastewater, but the microorganisms involved in this process and environment within the FGD effluent have not been well studied. This study identifies the microorganisms in selenate-enriched and selenite-enriched FGD effluent that can reduce selenium oxyanions to elemental selenium. Knowing more about these organisms can lead to an eventual optimization in the biological removal of selenium species in these wastewaters.

Sarkar, Preom↗
DOE OSTI2023

Rare earth enrichment process by contacting raw material with a base at specific pH values

Disclosed herein are methods for preparing a hydraulic pre-concentrate enriched in rare earth elements and critical minerals, the method comprising: (a) contacting a raw material with a first base in an amount sufficient to adjust the pH to a value from about 4.0 to about 6.0, thereby forming a mixture comprising a first aqueous phase and a first solid concentrate; (b) separating the first aqueous phase from the first solid concentrate; (c) contacting the first aqueous phase with a second base in an amount sufficient to adjust the pH to a value from about 7.0 to about 9.0, thereby forming a mixture comprising a second aqueous phase and the hydraulic pre-concentrate; (d) removing the second aqueous phase and collecting the hydraulic pre-concentrate; wherein the raw material comprises rare earth elements; and wherein the hydraulic pre-concentrate is enriched in rare earth elements.

Ziemkiewicz, Paul F.↗
DOE OSTI2024

High-Speed OH-PLIF Diagnostics of Flame Flashback in Low Swirl Hydrogen-Enriched Flames

Hydrogen-enrichment is a potential strategy to reduce carbon emissions for the development of next-generation lean premixed combustors. However, the higher flashback propensity with hydrogen enrichment owing to the faster chemical kinetics is one of the key challenges in existing natural gas turbine combustors. The fundamentals of stabilized flame dynamics and flashback events in a premixed low swirl burner (LSB) configuration are visualized experimentally using nanosecond (ns)-based high-speed hydroxyl radical planar laser-induced fluorescence (OH-PLIF). Swirlers with two different turning angles, 26 and 33 and three different center-body hole diameters, 1.08, 1.12 and 1.16 mm with measured swirl numbers varying from 0.43 to 0.49 are tested in this study at atmospheric temperature and pressure inlet conditions. The inlet conditions are varied systematically with respect to pre-mixer velocities (V), flame equivalence ratio (ϕ) and hydrogen content (X_(H_2 )) for 50%-90% (H2, by mole) H2/ CH4 reactant mixtures and OH-PLIF characterization is performed in a stable flame at 20 kHz repetition rate. An increase in flashback propensity is observed with increasing ϕ, X_(H_2 ) and decreasing V via detailed investigation of flame lift-off length (L) above the burner rim.

Parajuli, Pradeep↗
DOE OSTI2024

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗
DOE OSTI2024

Stabilizing high-Ni cathodes with gradient surface Ti-enrichment

High-Ni cathodes are being intensely pursued worldwide for electric vehicles and other energy-dense applications due to their high capacity and low cost. However, structural instabilities during electrochemical cycling and when subjected to thermal treatment have been the major issues hindering their practical deployment. We here report a rational design of coating-integrated-into-synthesis protocol for fabricating surface Ti-enriched LiNi 0.8 Mn 0.1 Co 0.1 O 2 (NMC811#Ti) material. The coating to intermediates is crucial to obtain high structural ordering, both in the bulk and surface of high-Ni cathodes, and the Ti substitute has a unique tri-valence (Ti 3+ ) in a gradient surface distribution. Further, the simulations of projected density of states in the atomistic understanding further certify significantly enhanced stability of lattice oxygen for the NMC811 through such a Ti 3+ -based structure reinforcement. Consequently, the NMC811#Ti cathode delivers a high capacity up to 200mAhg -1 at 0.1 C, along with superior stabilities during air-storage and thermal treatment (up to 297°C at the fully charged state under differential scanning calorimetric measurements). The corresponding NMC811#Ti||graphite full cell exhibits a desired 83.6% capacity retention after 1000 cycles at 0.5 C in a voltage range of 2.8–4.3V. This work demonstrates a delicate surface reinforcement to stabilize high-Ni cathodes for long-life and safe lithium-ion batteries.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Glass formulation and lab-scale testing of glasses designed for in-can Melter and in-container Vitrification of high-assay low-enriched uranium aqueous polishing Raffinate waste

Glasses were designed for processing a nuclear waste from aqueous polishing of high-assay low-enriched uranium using either In-Can Melter (ICM) or GeoMelt® In-Container Vitrification™ (ICV) technologies, which operate at temperatures of T p ≤ 1100 °C and T p ≤ 1450 °C, respectively. Due to the different operating conditions, the melt and glass properties were optimized differently for each technology. Each glass was designed to optimize for maximum waste loading while simultaneously satisfying processing (e.g., crystallization, viscosity, and conductivity) and product quality (e.g., durability, hazard characteristic, and crystal content) constraints. Here, the raffinate waste contains high nitric acid (4 M) and low total solids (9 g·L -1 ) concentrations. Feed preparation processes were tested to facilitate concentration and nitrate destruction/removal while controlling redox of the melter feed. A successful feed process including sugar addition and spray-drying was performed to generate an adequate melter feed for ICV processing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Benchmarking fast-neutron response of 7 Li-enriched C 7 LYC scintillators: Simulations and experiment

New insights into the fast neutron response of 7 Li-enriched C 7 LYC scintillators have been obtained through detailed Geant4 simulations of 35 Cl(n,p) and 35 Cl(n, α) reactions that are primarily responsible for the fast neutron response of C 7 LYC. The light output of a 1” x1” C 7 LYC detector was simulated using the Birks equation for light quenching in scintillators. The simulations were benchmarked against neutron scattering data from a 12 C target obtained with a white neutron beam at the Los Alamos LANSCE facility and fourteen 1”x1” C 7 LYC detectors. Incident neutron energies extracted from time-of-flight measurements, and appropriate recoil corrections for the different detector angles were used to map the response of C 7 LYC to mono-energetic neutrons. Resonances in the 35 Cl(n,p) reaction cross-section below 3 MeV incident neutron energy were directly observed in C 7 LYC. The experimental data and simulations are compared with ENDF databases, as well as recent direct cross-section measurements of neutrons scattered from a 35 Cl target. In conclusion, the results suggest a scrutiny of the 35 Cl(n,p) and 35 Cl(n, α) cross sections at low energy in the ENDF library.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Design of a High-Assay Low-Enriched Uranium Tri-Structural Isotropic Critical Experiment for Advanced Reactor Validation

High-assay low-enriched uranium (HALEU) fuel is a key component of many small modular reactor designs. Critical experiments are an important way to understand the neutronic performance of systems by obtaining nuclear data validations through measurements. Data reduce uncertainty and risk by showing that systems respond as predicted to changes such as temperature, subsequently advancing the overall technology readiness level of the materials within. Numerous critical experiments have been performed at the National Criticality Experiments Research Center (NCERC) operated by Los Alamos National Laboratory at the Nevada National Security Site since it became operational in 2011. However, the first experiment with HALEU fuel did not occur until 2024. Through extensive engineering, the experiment described in this paper was successfully designed and executed for the Comet vertical lift assembly at NCERC to perform measurements with HALEU tri-structural isotropic fuel that will assist in validation of nuclear data and computational modeling of small modular reactors for years to come.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Advanced High-Temperature Sodium-Cooled Thermal Reactors Using Less Than 10% Enriched UO 2 Fuel

Here, this paper presents a 1200-MW(thermal) advanced sodium-cooled thermal reactor concept that uses online refueling of 3.5% to 9.95% enriched UO2 fuel pin bundles; uses either graphite or beryllium oxide (BeO) as a neutron moderator; reaches outlet temperatures of 650°C enabling a thermal efficiency of at least 45%; has a high specific power of 133 W/g U; has average power densities of 16.4 and 43.2 W/cm 3 with graphite and BeO, respectively; reaches an average discharge burnup of 100 MWd/kg U; and generates 52% less spent fuel volume, 28% less fission products, and 47% to 64% less transuranics than a typical large pressurized water reactor for the same amount of electricity produced.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

The response of coarse root biomass to long-term CO 2 enrichment and nitrogen application in a maturing Pinus taeda stand with a large broadleaved component

Elevated atmospheric CO 2 (eCO 2 ) typically increases aboveground growth in both growth chamber and free-air carbon enrichment (FACE) studies. We report on the impacts of eCO 2 and nitrogen amendment on coarse root biomass and net primary productivity (NPP) at the Duke FACE study, where half of the eight plots in a 30-year-old loblolly pine (Pinus taeda, L.) plantation, including competing naturally regenerated broadleaved species, were subjected to eCO 2 (ambient, aCO 2 plus 200 ppm) for 15–17 years, combined with annual nitrogen amendments (11.2 g N m -2 ) for 6 years. Allometric equations were developed following harvest to estimate coarse root (>2 mm diameter) biomass. Pine root biomass under eCO 2 increased 32%, 1.80 kg m -2 above the 5.66 kg m -2 observed in aCO 2 , largely accumulating in the top 30 cm of soil. In contrast, eCO2 increased broadleaved root biomass more than twofold (aCO 2 : 0.81, eCO 2 : 2.07 kg m -2 ), primarily accumulating in the 30–60 cm soil depth. Combined, pine and broadleaved root biomass increased 3.08 kg m -2 over aCO 2 of 6.46 kg m -2 , a 48% increase. Elevated CO 2 did not increase pine root:shoot ratio (average 0.24) but increased the ratio from 0.57 to 1.12 in broadleaved species. Averaged over the study (1997–2010), eCO 2 increased pine, broadleaved and total coarse root NPP by 49%, 373% and 86% respectively. Nitrogen amendment had smaller effects on any component, singly or interacting with eCO 2 . A sustained increase in root NPP under eCO 2 over the study period indicates that soil nutrients were sufficient to maintain root growth response to eCO 2 . These responses must be considered in computing coarse root carbon sequestration of the extensive southern pine and similar forests, and in modelling the responses of coarse root biomass of pine–broadleaved forests to CO 2 concentration over a range of soil N availability.

9 allometry↗
DOE OSTI2022

Reactor physics benchmark of Westinghouse PWR core design suite for high burnup/high enrichment fuel - Part 1: Pin and lattice

Westinghouse has performed a comprehensive set of code-to-code benchmark comparisons to corroborate application of its state-of-the-art NEXUS-based code suite to high burnup and high enrichment (HB/HE) fueled PWR core design and reload analysis. This paper focuses on analysis of the predictions for pin and lattice designs representative of HB/HE core reloads from the Westinghouse advanced lattice code PARAGON2, the Monte-Carlo code SERPENT2 and MPACT, the deterministic neutronic component of VERA. A companion paper presented at this conference analyzes results obtained with the Westinghouse ANC9 core simulator, using lattice data provided by PARAGON2 with the NEXUS cross-section methodology, and MPACT for 2D and 3D core configurations incorporating HB/HE fuel reloads to achieve cycle length extension to 24-month in high-power density PWRs. The results of the pin and lattice benchmark presented in this paper show that the neutronic predictions from PARAGON2 for HB/HE fueled PWRs are in remarkable agreement with predictions from both SERPENT2 and MPACT including for depletion to high fuel burnup values, and employing a set of challenging lattices which feature combined use of burnable absorbers to obtain the required reactivity hold-down to support PWR transition to HB/HE fuel and 24-month operational cycles. As corroborated by this analysis and the analysis contained in the companion paper, the Westinghouse NEXUS-based PWR core analysis package can be confidently applied by industry for design analysis of PWR core reloads adopting HB/HE fuel and properly support the industry commercialization plans for in-reactor deployment. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Validation of MCNP Critical Benchmark Models of Moderated Highly Enriched Uranium Slabs

A new centralized repository of high-quality MCNP models of critical benchmark experiments is currently under development at Los Alamos National Laboratory (LANL). The benchmark experiments are described in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and the initial set of benchmark models are derived from the Whisper Suite provided with MCNP6.2. This effort is a collaboration among the Nuclear Criticality Safety, Nuclear Data, and Monte Carlo code development/application organizations at LANL. The goal is to build a single LANL benchmark collection that is up to date with the latest ICSBEP revision, has a formal review and revision process, is contained in an open-source repository, and utilizes new Python tools for improved input and output file review. This paper describes the validation of the models associated with HEU-MET-FAST-007, “Uranium Metal Slabs Moderated with Polyethylene, Plexiglas, and Teflon”. The Monte Carlo n-Particle (MCNP) models were compared to the second revision of HEU-MET-FAST-007. The experiment considered critical configurations of highly enriched uranium (HEU) metal slabs and various moderators in 43 unique cases. The slabs of uranium were separated by layers of moderating material, forming an assembly. The assembly was separated in two halves with one half placed on a stationary table and the other attached to a moveable table that could be raised and lowered via pulleys connected to the ceiling of the shielded room. To reach criticality, the two halves were brought together with a negligible gap between the assemblies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

TEX-Hf: Integral Experiment Execution of Thermal/Epithermal eXperiments using Highly Enriched Uranium with Polyethylene and Hafnium (IER-532 CED-3b Report)

This report documents the experimental configurations and measurements for IER-532, Thermal/Epithermal eXperiments (TEX) with highly enriched uranium (HEU) fuel and interstitial hafnium (Hf), moderated and reflected by polyethylene. TEX-Hf is a variation of and based on the TEX-HEU (IER-297) design, with the inclusion of hafnium. These configurations provide integral experiments for validation of hafnium in the thermal, intermediate, and fast neutron energy regimes by maximizing the sensitivity in k eff to the hafnium isotope cross sections. The experiment campaign was completed over seven weeks during the end of FY22 and beginning of FY23 at the National Critical Experiments Research Center at the Nevada National Security Site. The campaign produced seven experimental configurations, four reproducibility measurements, and many additional dimensional measurements that will be of use to the future benchmark evaluation of this experiment and other experiments using the same HEU fuel. Table 1 summarizes the TEX-Hf experimental configurations, including their physical parameters, calculated fission fractions, and estimated excess reactivities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Delayed Critical Highly Enriched Uranium Metal Cylinders with Thin Graphite Top and Bottom Reflectors

Nine 7, 11, and 15 in. diam highly enriched uranium (HEU, 93.15 wt % 235 U) metal cylinders were assembled on the vertical assembly machine in the Oak Ridge Critical Experiments Facility (ORCEF) and had 1, 2, or 3 in. thick HLM graphite reflectors on the top and bottom. The experiments, which were performed between April 3, 1970, and February 18, 1971, used 23 operational days at ORCEF. Before those experiments were carried out, unreflected and unmoderated, graphite- and polyethylene-reflected, and polyethylene-moderated HEU metal cylinders had been assembled to obtain delayed criticality at ORCEF in the 1960s and reported by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) at the Nuclear Energy Agency (NEA)*. The data from the nine critical experiments are acceptable for use as criticality safety benchmark experiments for the NEA’s ICSBEP once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with HEU metal at ORCEF, the uncertainties in k eff are expected to be as low as ±0.0004.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗