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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 1,261 records · Page 70
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Continuous partial oxidation of methane to methanol over Cu-SSZ-39 catalysts

Here, the direct conversion of methane to methanol using molecular oxygen as the oxidant is a grand challenge in the chemical sciences. Here we describe our efforts in using copper-exchanged SSZ-39 for this chemistry and show it is a promising material for this reaction. Testing in a continuous flow reactor at 225 °C, Cu-SSZ-39 showed superior activity for methane partial oxidation compared to Cu-SSZ-13 (site time yields of 26.9 ± 0.8 mol CH3OH ·mol Cu -1 ·h -1 *10 -3 versus 13.2 mol CH3OH ·mol Cu -1 ·h -1 *10 -3 respectively). A clear maximum in methanol productivity, both in terms of a mass catalyst basis and copper atom basis is observed for Cu-SSZ-39 at a Cu/Al ~ 0.2. Running this reaction in the absence of oxygen over samples with a Cu/Al > 0.1 the reactivity is markedly lower than that in the presence of oxygen unlike what is observed in SSZ-13.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Transport and kinetics properties of LaF 3 in FLiNaK molten salt determined by electrochemical methods

It is found that lanthanum species can exist in the form of LaF 6 3- in FLiNaK (46.5 LiF-11.5 NaF-42 KF, mol%) molten salt in previous studies. The understanding of transport and reaction kinetics parameters of LaF 6 3- ions in FLiNaK molten salt is essential for the electrochemical separation of lanthanum from FLiNaK melt to achieve the coolant clean-up in fluoride salt cooled high-temperature reactor (FHR). Through conducting chronopotentiometry tests, the diffusion coefficients of LaF 6 3- ions in FLiNaK molten salt at the temperature range of 923 K–1023 K were determined. By conducting potentiodynamic polarization measurements at different concentrations and temperatures, the exchange current density, reaction rate constant, and charge transfer coefficient were obtained. The experiment data of potentiodynamic polarization were analyzed using a non-simplified electrode kinetics equation which incorporates both mass transfer and reaction kinetics.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

In this paper, we present post-irradiation examination results on two type of annular mixed oxide fuel pins irradiated in the Fast Flux Test Facility (FFTF) sodium cooled reactor to an average burnup between 4% and 5% fission of initial heavy atom (FIMA). The pins differed only from the initial Pu content, which was 22 wt% and 26 wt%, respectively. The overall performance of the pins was excellent, in line with previous historical results. The pins with higher Pu content experienced higher irradiation temperatures which influenced the fission gas release, fuel swelling, and Cs distribution compared to the other pins. All the post-irradiation examinations results are discussed against the irradiation parameters. In particular, the pins with higher initial Pu content, i.e., 26 wt%, experienced higher power that resulted in enhanced fission gas release compared to the other two pins with 22 wt% initial Pu content. For the pins with higher fission gas release, onset of Cs redistribution was observed. The two pins that had lower initial Pu content and burnup showed a Cs axial distribution similar to the as-produced one.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

A comparison of wrought and powder metallurgical FeCrAl claddings under simulated LWR accident transients

Iron-chromium-aluminum (FeCrAl) alloys are potential accident tolerant fuel (ATF) cladding candidates for light-water reactors but are difficult to fabricate as thin-walled tubes via conventional cast-and-wrought routes. Powder metallurgy (PM) offers a manufacturing alternative with improved compositional control, but its transient accident performance has not been directly benchmarked against wrought variants. This study evaluates the burst behavior of commercially developed PM-processed FeCrAl alloys, PM-C26M (Fe-12Cr-6Al-2Mo) and the high precipitate density FA-SMT (Fe-22Cr-5Al-3Mo), under simulated light-water reactor accident transient conditions. Burst testing was conducted using the Severe Accident Test Station with heating rates of 5 °C/s and 50 °C/s and internal pressures ranging from 25 MPa to 100 MPa. PM-C26M reproduced wrought C26M burst behavior within 7–37 °C across the stress range, indicating that PM processing does not compromise transient strength. FA-SMT exhibited markedly higher burst temperatures and reduced heating-rate sensitivity, consistent with its engineered precipitate strengthening. FA-SMT rupture exhibited axial "unzipping" rather than the lateral tearing characteristic of PM- and wrought C26M. Post-test EBSD and fractography indicate that this behavior is strongly correlated with strain-gated intergranular void nucleation associated with the dense precipitate architecture of FA-SMT, a response absent in the comparatively clean PM-C26M matrix and consistent with rupture morphologies reported for oxide-dispersion strengthened (ODS) FeCrAl of similar base-matrix chemistry to PM-C26M. These findings highlight the potential of powder metallurgy as a viable fabrication route for ATF claddings from an accident performance standpoint.

Bell, Sam [ORNL] (ORCID:0000000251905657)↗
DOE OSTI2025

Mechanistic Multiscale Uncertainty Propagation in Support of Accelerated Fuel Qualification

Taking a nuclear fuel concept through the research, development, and qualification stages has historically taken on the order of 20 to 25 years because of extensive irradiation tests required for a variety of conditions. The concept of accelerated fuel qualification (AFQ) has been proposed to increase the innovation pace for nuclear fuels. The goal of AFQ is not to replace the traditional qualification approach but rather to reduce the total number of experiments required to ensure approval from the regulatory authority. Of the many AFQ approaches being explored, advanced modeling—and, in particular, mechanistic modeling—is in a uniquely cross-cutting position to reduce the number of required integral tests through the inclusion of separate-effects testing, while helping to extrapolate reactor performance during rare events. We make the case that propagation of uncertainty through various computational length scales helps contextualize mechanistic modeling. We will utilize UO 2 fission gas diffusion predictions from the atomistically informed cluster dynamics code Centipede to inform fuel performance rodlet simulations using the BISON finite element code as the metric for showing how multiscale mechanistic uncertainty quantification can help reduce uncertainty in fuel performance. In conclusion, by quantifying uncertainty and its reduction through multiscale modeling, the qualification process may be accelerated through the reduction of costly irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microreactor Automated Control System Test Bed Digital Architecture for Real-Time, Hardware-in-the-Loop Simulation

This work describes progress made towards the development of a real-time hardware-in-the-loop (HIL) test bed for non-nuclear testing of microreactor control schemes and failure modes. Non-nuclear testing is a crucial step in developing robust control algorithms for managing microreactor dynamics. The creation of an HIL simulation harnesses the realistic dynamics of physical analogue systems while additionally considering the challenges of variable communication delay. This collaborative effort between Oak Ridge National Laboratory and Idaho National Laboratory has resulted in a LabVIEW-based gRPC communication protocol which couples a TRANSFORM Modelica simulation of nuclear components to the ViBRANT physical hardware for realistic feedback and visual representation of control action in real time. A modular python client structure is developed to manage FMU-based Modelica simulation and real-time gRPC communication. HIL testing suggests that the modeled reactor with natural convection molten salt loop coolant configuration responds well to PID control of drum positioning for modulation of reactor core power, however, future efforts will be made to explore the added thermal inertial delay of system level control and downstream demand changes. Development of this platform with a generalized methodology provides a foundation for exploring a variety of reactor configurations and failure modes in rapid order to provide insight into the most effective avenues of study for further research and development.

McConnell, Jono [ORNL] (ORCID:0000000238984741)↗
DOE OSTI2020

SCALE Code System

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Subcritical Multiplication Ex-Core Calculations with Shift

This report details the development and testing of Virtual Environment for Reactor Applications (VERA) as applied to full core subcritical multiplication ex-core calculations with Shift. The limiting factor in being able to run these detector calculations with Monte Carlo is the amount of computational memory used by a fully-loaded full core problem. The main development features added to VERA in this effort were enabling source biasing and separable weight windows and integrating multithreading in Shift. With these capabilities, a subcritical multiplication detector response calculation on a full core Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was successfully run on a moderately sized computing cluster.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Time Dependence of Methane Generation Rates from Saltstone

This report documents the tests performed to address the requirements listed in the Technical Request, which resulted in a Task Technical & Quality Assurance Plan (TTQAP) and a Run Plan to obtain the required information. The research expands previous scoping work to better understand methane generation in the Saltstone Disposal Facility (SDF). That scoping work demonstrated that methane is produced when sulfide-containing material is added to methylmercury (MeHg), which is found in the SDF Tank 50 feed waste. When Tank 50 waste is combined with grout premix solids dry feed ingredients of cement, slag cement, and fly ash, sulfides are released from the slag cement component of the premix. The prior testing results showed that while the concentration of total mercury was not affected in the pH 14 waste, the MeHg likely reacted with the sulfide contained in the slag. The scoping tests were performed in sealed reactors and thus, methane generation was not measured continuously; rather, it was measured after one hour and after 24 hours. After 24 hours, a large percentage of the MeHg was converted to methane, but after 1 hour only a small percentage of methane was released. It is noteworthy that the methane generation rate (MGR) was greater at the one-hour measurement. The data compiled from the scoping work needed to be expanded to better understand the generation of methane released from grout mixed with Tank 50 waste. This expanded study developed a more extensive knowledge of how and when methane is generated to inform the Documented Safety Analysis (DSA) of the SDF. As such, employing 8 sealed reactors, the methane generation time periods were extended to 2, 4, 6, 18, 30, 48, 72, and 96 hours (4 days). Results are documented herein to determine the capacity for MeHg, contained within Tank 50 salt solution (T50SS) simulant, to produce methane when mixed with saltstone dry feed ingredients, i.e., individual components (cement, slag, and fly ash) or a combination of those components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Implementation of the Windowed Multipole Method in Shift

The windowed multipole (WMP) method has been implemented in the Shift Monte Carlo (MC) radiation transport code with support for both CPU and GPU execution. With this method, small WMP data libraries (~100 MB) can be used to accurately Doppler broaden cross sections to arbitrary temperatures “on the fly” during an MC simulation. This approach yields significant memory savings relative to traditional methods, making it ideal for high-fidelity analysis such as coupled multiphysics simulations. This document provides the exact forms of the WMP equations used by Shift, as well as a detailed description of the structure of WMP HDF5 data files provided by the Massachusetts Institute of Technology (MIT). The Shift implementation has been validated against the OpenMC radiation transport code, with excellent agreement demonstrated for 70 nuclides across an operative range of temperatures. CPU and GPU performance testing using a small module reactor (SMR) problem demonstrated that this method decreases the neutron tracking rate by a factor of ~2 on the Summitdev machine. A new set of WMP data being developed in-house will employ novel methods to improve tracking rates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Aluminization Coatings and Glass Seals for High Temperature System: CRADA 536 [Abstract Only]

The proposed work will demonstrate the suitability of PNNL Reactive Air Aluminization (RAA) coatings (US Patent Numbers 10,577,694, 10,378,094, and 9,481,923) for use in high-temperature systems. In particular, the Participant and Contractor will explore applicability of these coatings in Participant’s high-temperature hydrogen reactors where they would be used to prepare the sealing surfaces of the reactor technology. Utility Global will also instruct PNNL regarding their needs relating to the further development of glass seal technologies that are suitable for use in their systems. This effort will entail alkaline earth-based silicate glass development and characterization; evaluation and characterization of reliability and compatibility of the coatings with glass seals, working environments, and substrates; performance demonstrations of representative cells and intermediate cells sealed using the glass seals and/or coatings of interest; and incorporation of the aluminization coatings and glass seals into the reactor for full-scale testing.

08 HYDROGEN↗
DOE OSTI2022

Thermal Hydraulic Experimental Test Article - Fiscal Year 2022 Final Report

The Thermal Hydraulic Experimental Test Article (THETA) is a METL vessel experiment designed for testing and validating sodium fast reactor components and phenomena. THETA has been scaled using a non-dimensional Richardson number approach to represent temperature distributions during nominal and loss of flow conditions in a sodium fast reactor (SFR), this analysis was detailed in the THETA FY19 report. The facility is being constructed with versatility in mind, allowing for the installation of various immersion heaters, heat pipes, and heat exchangers without significant facility modification. THETA was designed in collaboration with systems code experts to inform the geometry and sensor placement to acquire the highest value code validation data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

SULI Report - Development of a Molten Salt Circulation Loop for in-situ Spectroscopy

This project supports the development of real time optical monitoring capabilities for molten salt reactor (MSR) environments by designing, testing, and refining a molten salt circulation loop suitable for combined laser induced breakdown spectroscopy (LIBS) and ultraviolet visible (UV Vis) absorption measurements. Online spectroscopic monitoring is increasingly important for nuclear safeguards, corrosion tracking, and material accountancy, yet MSR process fluids present substantial challenges due to their chemical complexity and hazards such as high temperatures and radiation. To address these needs, this work focuses on Phase I, the development of a room temperature aqueous circulation loop that serves as a surrogate platform for evaluating flow behavior, optical access, and component performance prior to high temperature salt operation. Initial testing identified several practical issues—including leaks, obstructions, and two-phase flow through the absorption cell—that were systematically resolved through hardware replacement, flow path redesign, and venturi pressure optimization. Relocating the flow cell upstream of the primary venturi enabled periods of stable single-phase flow, demonstrating the feasibility of integrating optical diagnostics into a circulation system. The results of Phase I provide essential design insight for Phase II, which will incorporate furnace compatible materials and LiCl KCl eutectic salt. Completion of the molten salt system will deliver a reusable testbed for evaluating multimodal spectroscopic techniques, advancing nondestructive, real time monitoring tools for future MSR and nuclear fuel cycle applications.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2025

A High-Fidelity Model of the Peach Bottom 2 Turbine-Trip Benchmark Using VERA

This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type.

BWR↗
DOE OSTI2020

Project Methods to Enhance Wellbore Cement Integrity with Microbially-Induced Calcite Precipitation (MICP) (Final Scientific/Technical Report)

The goal of this project was to develop improved methods for sealing compromised wellbore cement in leaking oil and gas wells, thereby reducing the risk of unwanted upward fluid migration. Novel methods for improving wellbore integrity, such as microbially induced calcite precipitation (MICP), can reduce leakage potential, improve the safety of fossil fuel extraction, improve the public perception of hydraulic fracturing, and promote environmentally-prudent unconventional oil and gas development. Microbes, with the urease enzyme, can catalyze the chemical reaction of urea hydrolysis to induce the precipitation of calcium carbonate which can be used as a cementitious material to seal leakage pathways. In this project, methods to promote robust bio-composite cementitious materials were designed and tested in the laboratory. Scale-up of those methods were tested in meso-scale reactor systems and in field applications. In this report, in Section One, we describe laboratory efforts to develop injection strategies to promote precipitation in wellbore analogs and determine the strength of the bio-composite cements as compared to fine cement. In Section Two, we describe the efforts to scale up the work and study the use of materials that can be used in field application, for example exploring the use of calcium chloride ice melt or urea fertilizer as source chemicals. In Section Three, the three field trials (methods and results) performed as part of the project are described and summarized. At the end of the report is a comprehensive summary and conclusion section which highlights the key findings of the project. The work performed during this project significantly advanced the technology readiness level (TRL) of the MICP wellbore sealing strategy.

02 PETROLEUM↗
DOE OSTI2020

Methods to Enhance Wellbore Cement Integrity with Microbially-Induced Calcite Precipitation (MICP) (Final Scientific/ Technical Report)

The goal of this project was to develop improved methods for sealing compromised wellbore cement in leaking oil and gas wells, thereby reducing the risk of unwanted upward fluid migration. Novel methods for improving wellbore integrity, such as microbially induced calcite precipitation (MICP), can reduce leakage potential, improve the safety of fossil fuel extraction, improve the public perception of hydraulic fracturing, and promote environmentally-prudent unconventional oil and gas development. Microbes, with the urease enzyme, can catalyze the chemical reaction of urea hydrolysis to induce the precipitation of calcium carbonate which can be used as a cementitious material to seal leakage pathways. In this project, methods to promote robust bio-composite cementitious materials were designed and tested in the laboratory. Scale-up of those methods were tested in meso-scale reactor systems and in field applications. In this report, in Section One, we describe laboratory efforts to develop injection strategies to promote precipitation in wellbore analogs and determine the strength of the bio-composite cements as compared to fine cement. In Section Two, we describe the efforts to scale up the work and study the use of materials that can be used in field application, for example exploring the use of calcium chloride ice melt or urea fertilizer as source chemicals. In Section Three, the three field trials (methods and results) performed as part of the project are described and summarized. At the end of the report is a comprehensive summary and conclusion section which highlights the key findings of the project. The work performed during this project significantly advanced the technology readiness level (TRL) of the MICP wellbore sealing strategy.

03 NATURAL GAS↗
DOE OSTI2024

Fiber-Based Distributed Hydrogen Measurement at High Temperatures

In this work, we discuss the implementation of evanescent field optical fiber sensors for distributed gas sensing of H2, focusing on results using a Ni / Gd-doped CeO2 nanocomposite sensing material for detection of low levels of H2 (0-4%) at 700 °C. This approach utilizes sensors prepared using a low-cost, wet chemical deposition process, in conjunction with a custom-built interrogator system leveraging optical time domain reflectometry (OTDR). Using a specially designed dual-gas flow reactor system, the sensor is tested by establishing a controlled equilibrium gradient of gas concentration. Initial results are shown using silica fiber, but provides a pathway for utilization with high temperature stable single crystal optical fiber for operation at higher temperatures and higher levels of H2 relevant for solid oxide fuel cell (SOFC) operating conditions.

Wuenschell, Jeffrey↗