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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Application of Proliferation Resistance Optimization (PRO-X) Methodology to a Generic Research Reactor

The Proliferation Resistance Optimization Program (PRO-X) has been established by the NNSA to provide a framework for evaluating and integrating proliferation resistance into nuclear reactor system (core, fuel, and auxiliary facilities) designs that also maintain the safety and peaceful use missions of those systems. The research reactor (PRO-RR) area is the component of PRO-X that supports the program objectives by applying state-of-the-art analysis methods to research reactor systems. This report details the results of using a defined set of analytical tools to evaluate the neutronics, thermal hydraulics and proliferation risk characteristics of a set of parametric cores based on a generic 10 MW materials test reactor that uses a plate-type low enriched uranium (LEU) fuel. The analysis shows that by suitable adjustment of the core size, reflector configuration and power level, mission performance and safety margins can be maintained or improved while reducing the potential for production of special nuclear material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Novel Technology of Non-Contact Real-Time Radiation Damage Sensor for High Power Targets

High-power proton beams planned for forthcoming long-baseline neutrino experiments will subject solid targets to unprecedented radiation damage, threatening reliability and increasing costs. To address this challenge, we are building a real-time, non-contact sensor that tracks damage by measuring broadband laser reflectivity changes from the target surface. A low-power super-continuum Class 3B laser illuminates the sample inside a vacuum test chamber while a high-resolution fiber-coupled spectrometer captures S- and P-polarized light; spectral shifts reveal defect-driven variations in optical constants. My internship goal is to design, build, and commission this prototype by implementing a laser-safety interlock and light shield, integrating remote-operation cameras, and fabricating modular 3-D-printed mounts that enable tool-free swaps without disturbing alignment. The laser, spectrometer, and vacuum chamber have been delivered; interlock hardware, cameras, and mounts are in final assembly, and leak testing of the chamber is underway. Upcoming work will focus on initial calibration of the sensor and executing first beam-irradiation studies.

Pumarino, Rafael↗
DOE OSTI2021

Quantitative resilience evaluation on recovery from emergency situations in nuclear power plants

Here, this paper mainly introduces how to develop a resilience evaluation model by quantifying the relationship between resilience and resilience components in regard to recovering from emergency situations in nuclear power plants (NPPs). It is an extension of the author’s previous researches. To develop the model, the first step was to analyze event reports published in Republic of Korea in terms of resilience factors from different safety perspectives. Second, statistical methods such as factor analysis, principal component analysis, and multivariate regression analysis were applied to the dataset to identify relations among resilience factors and develop a quantitative resilience evaluation model for recovery from emergency situations in NPPs. Third, the model was validated using the recent event report data released in Korea. Lastly, we present a discussion on how to apply the quantitative resilience evaluation model to results from stress tests performed in Korea.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Continuous Integration, In-Code Documentation, and Automation for Nuclear Quality Assurance Conformance

The Multiphysics Object Oriented Simulation Environment (MOOSE) is an open-source, finite element framework for solving highly coupled sets of nonlinear equations. The development of the framework and applications occurs concurrently using an agile, continuous-integration software package. Included in the framework is an in-code, extensible documentation system. Using these two tools in union with the repository management tools GitHub and GitLab, a software quality plan was created and followed such that MOOSE and a MOOSE-based application (BISON) have been shown to meet the American Society of Mechanical Engineers’ Nuclear Quality Assurance-1 standard. The approach relies heavily on automation for both testing and documentation. The resulting effort demonstrates that a rigorous software quality plan may be implemented that incurs a minimal impact on day-to-day development of the software, satisfying the stringent guidelines necessary to operate the software in a safety function within a nuclear facility.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

COVID-19 Exposure Assessment Tool (CEAT): Exposure quantification based on ventilation, infection prevalence, group characteristics, and behavior

The coronavirus disease 2019 (COVID-19) Exposure Assessment Tool (CEAT) allows users to compare respiratory relative risk to severe acute respiratory syndrome coronavirus 2 (SARS-CoV-2) for various scenarios, providing understanding of how combinations of protective measures affect risk. CEAT incorporates mechanistic, stochastic, and epidemiological factors including the (i) emission rate of virus, (ii) viral aerosol degradation and removal, (iii) duration of activity/exposure, (iv) inhalation rates, (v) ventilation rates (indoors/outdoors), (vi) volume of indoor space, (vii) filtration, (viii) mask use and effectiveness, (ix) distance between people (taking into account both near-field and far-field effects of proximity), (x) group size, (xi) current infection rates by variant, (xii) prevalence of infection and immunity in the community, (xiii) vaccination rates, and (xiv) implementation of COVID-19 testing procedures. CEAT applied to published studies of COVID-19 transmission events demonstrates the model’s accuracy. We also show how health and safety professionals at NASA Ames Research Center used CEAT to manage potential risks posed by SARS-CoV-2 exposures.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Nimble Feedthrough Qualification - 125% High Explosive Overpressure Test Plan RevA

The primary purpose of this High Explosive (HE) Over-Pressure Test (OPT) is to qualify top cover diagnostic feedthroughs that will be used on LLNL Nimble Subcritical Experiment (SCE) Series designs per experimental design verification requirements specified in ASME Boiler and Pressure Vessel Code Case 2564, Impulsively-Loaded Pressure Vessels, Section VIII, Division 3; and to satisfy the over-test requirement of DOE-STD-1212. The diagnostic feedthroughs are part of the Vessel Confinement System (VCS), which is credited as a Safety Significant Design Feature per the U1a Facility Documented Safety Analysis (DSA). The OPT will be conducted at the LANL Area 1, R306 Firing Site (TA-15-R306) in a 3-foot diameter VCS depicted in Figure 1.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2022

A Review of the Literature on the Radiolytic Stability of the Next-Generation Solvent

The Original Caustic-Side Solvent Extraction (CSSX) solvent (based on the BOBCalixC6 extractant) currently in use at the Salt Waste Processing Facility (SWPF) is set to be replaced by the Next-Generation Solvent (NGS, based on the MaxCalix extractant). Current SWPF flammability safety controls use G-values for volatile gases obtained from radiolysis studies with the Original CSSX solvent. A review of past irradiation tests of both solvent systems was performed to assess the radiolytic stability of NGS with respect to flammable gas generation from radiolysis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Numerical investigation of the AP1000 response following loss-of-coolant accident using PCTRAN and CFD to support R&D of SMRs

Small Modular Reactors (SMRs) present a promising solution for the future of sustainable energy, offering advantages such as reduced waste generation, advanced passive safety features, and potential cost efficiencies. Ongoing research focuses on the design and development of SMRs, addressing challenges through numerical simulations and experimental test data. Given that many next-generation reactors are first-of-a-kind (FOK), current operational reactors serve as valuable benchmarks for understanding various thermal-hydraulic phenomena during postulated design basis accidents (DBAs), including loss-of-coolant accidents (LOCA), main steam line breaks (MSLB), and steam generator tube ruptures (SGTR). This study employs the AP1000 model Personal Computer Transient Analyzer (PCTRAN) to simulate reactor responses to small-break LOCA scenarios, specifically analyzing break sizes of 2 and 10 inches. Additionally, ANSYS FLUENT software is utilized to assess containment responses to large-break LOCAs, focusing on the quantification of decay heat removal via natural convection. Verification of the PCTRAN results is achieved using Westinghouse data, with CFD results corroborating the findings. The outcomes demonstrate strong agreement with Westinghouse data, confirming the accuracy of the simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

PELICAN Status Report for Year 2021

The Pressure Drop Experimental Loop for Investigations of Core Assemblies in Advanced Nuclear Reactors, PELICAN, is a novel experimental facility designed and constructed to recreate the prototypic full scale hydraulic flow conditions for the Versatile Test Reactor (VTR) currently under development by the U.S. Department of Energy. To support VTR design efforts, PELICAN is used to replicate VTR’s primary heat transport system (PHTS) in the core assembly region. The measurement objectives for data generated from this test facility are to mimic the prototypic VTR assembly and to meet the validation needs for code calculations and simulations of the reference design for the VTR core. These objectives focus on the need to validate pressure drop results across the various segments of the fuel assembly, including the axial reflectors, fuel, and transition regions, as they relate directly to the pumping power and safety behavior of the reactor. In the following sections, the major milestones for the experimental work performed on PELICAN over the calendar year (CY) 2021 are presented. In the next section, we discuss the assembly and shakedown testing of the facility. In Section 3, an analysis of the errors in the major measurements is performed. In Section 4 and 5 we present the results from the first set of test articles placed inside the loop. The design of the remaining test articles as part of the full assembly are shown in Section 6 and the conclusion and path forward for the following year, CY22 is given in the final section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Urban working groups in the IAEA’s model testing programmes: overview from the MODARIA I and MODARIA II programmes

The IAEA’s model testing programmes have included a series of Working Groups concerned with modelling radioactive contamination in urban environments. These have included the Urban Working Group of Validation of Environmental Model Predictions (1988–1994), the Urban Remediation Working Group of Environmental Modelling for Radiation Safety (EMRAS) (2003–2007), the Urban Areas Working Group of EMRAS II (2009–2011), the Urban Environments Working Group of (Modelling and Data for Radiological Impact Assessments) MODARIA I (2013–2015), and most recently, the Urban Exposures Working Group of MODARIA II (2016–2019). The overarching objective of these Working Groups has been to test and improve the capabilities of computer models used to assess radioactive contamination in urban environments, including dispersion and deposition processes, short-term and long-term redistribution of contaminants following deposition events, and the effectiveness of various countermeasures and other protective actions, including remedial actions, in reducing contamination levels, human exposures, and doses to humans. This paper describes the exercises conducted during the MODARIA I and MODARIA II programmes. These exercises have included short-range and mid-range atmospheric dispersion exercises based on data from field tests or tracer studies, hypothetical urban dispersion exercises, and an exercise based on data collected after the Fukushima Daiichi accident. We report improvement of model capabilities will lead to improvements in assessing various contamination scenarios (real or hypothetical), and in turn, to improved decision-making and communication with the public following a nuclear or radiological emergency.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Applications (Acceptance Testing Report)

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal- cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over fifty years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. To support demonstration of software pedigree and confirm key functional requirements, this report has been generated to provide detailed verification of the software. Although SAS was developed to support the analysis of any liquid-metal-cooled nuclear reactor, the acceptance testing described in this document focuses on the verification of SAS capabilities as they relate to pool-type Sodium Fast Reactors (SFRs). This report includes documentation of the test problem definition, analytical solution(s), computational solution(s), acceptance criteria, comparisons of analytical/computational solutions, and determination of acceptance of the computational solution(s). Deviations from acceptance criteria are noted.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Landfilled coal ash for carbon dioxide capture and its potential as a geopolymer binder for hazardous waste remediation

In this report, the legacy of global fossil fuel usage is not just the increase in global warming due to CO 2 production, but also the billions of tons of toxic coal ash disposed in landfills. This investigation evaluates the merits of using high-calcium landfilled coal fly ash of different disposal durations and granulated blast furnace (GBF) slag as primary raw materials for production of a new class of geopolymer binder. Processing of landfilled coal fly ash, GBF slag and supplementary raw materials into a geopolymer binder was accomplished by a simple, scalable milling process in the absence of heating under a carbon dioxide environment. The resultant geopolymer cement binders were tested for carbon dioxide uptake, compressive strength, and heavy metal stabilization. The geopolymer binder demonstrated a desired balance of carbon dioxide uptake, strength and safety. These results also indicate that the performance of the geopolymer binder depends on the landfill disposal age of the coal fly ash.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2020

Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis

Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test (TREAT) facility at Idaho National Laboratory was developed. The experiment named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless steel heater rod submerged in a static water pool which is heated via the (n, a) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software DAKOTA to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Investigating the Effects of Lithium Deposition on the Abuse Response of Lithium-Ion Batteries

Li deposition at the graphitic anode is widely reported in literature as one of the leading causes of capacity fade in lithium-ion batteries (LIBs). Previous literature has linked Li deposition resulting from low-temperature ageing to diminished safety characteristics, however no current research has probed the effects of Li deposition on the abuse response of well-characterized cells. Using overtemperature testing, a relationship between increased concentrations of Li deposition and exacerbated abuse response in 1 Ah pouch cells has been established. A novel Li deposition technique is also investigated, where cells with n:p < 1 (anode-limiting) have been cycled at a high rate to exploit Li + diffusion limitations at the anode. Scanning Electron Microscopy of harvested anodes indicates substantial Li deposition in low n:p cells after 20 cycles, with intricate networks of Li(s) deposits which hinder Li + intercalation/deintercalation. Peak broadening and decreased amplitude of differential capacity plots further validates a loss of lithium inventory to Li + dissolution, and Powder X-ray Diffraction indicates Li + intercalation with staging in anode interstitial sites as the extent of Li deposition increases. A cradle-to-grave approach is leveraged on cell fabrication and testing to eliminate uncertainty involving the effects of cell additives on Li deposition and other degradation mechanisms.

25 ENERGY STORAGE↗
DOE OSTI2024

H2 Materials Compatibility of Low Cost, High Pressure, Polymer H2 Dispensing Hoses

NanoSonic was recently awarded a DOE Phase IIB SBIR program for the commercialization of an innovative metal-free polymer based H2 dispensing hose to make H2 an economically viable fuel alternative to gasoline. During the Phase I and Phase II base programs, NanoSonic's H2 hose demonstrated ultra-high hydrostatic burst strength values > 31,000 psi where failure occurred due to fitting slippage, rather than hose burst. Additionally, our hoses survived > 51,000 pressure impulse cycles at 12,000 psi over a thermal cycle of -40 °C to 85 °C. These hoses also failed due to fitting slippage. Thus, our Phase IIB program is centralized on the addition of a new fitting, polymer hose refinement, and validation of a complete hose and fitting system certified for use with H2. While NanoSonic has partnered with the National Renewable Energy Laboratory (NREL) to test our hose on their robotic H2 dispensing system, we are also seeking durability with the H2 environment at the molecular level. Here, NanoSonic proposes a CRADA with Pacific Northwest National Laboratory (PNNL), who has expertise in H2 polymer materials compatibility. Together, we will determine the lifetime of our H2 hose polymer and composite constituents via: 1) time-temperature superposition (TTS) studies via dynamic mechanical analysis (DMA) under H2, 2) friction and wear resistance under in situ H2 tribometry, and 3) multi-axis strain testing under cryogenic conditions. PNNL has the unique test equipment essential to gain this type of insight and lifetime prediction data. Importantly, the work conducted under the proposed CRADA will increase the safety and reliability of our H2 hoses while expanding the market for use of our low H2 permeation polymers and durable cryogenic composites to realize the H2@Scale objectives to reduce the cost of H2.

08 HYDROGEN↗
DOE OSTI2025

Hazard Detection Detector Cards

This report presents a comprehensive summary of five advanced anomaly detection tools developed and deployed by Oak Ridge National Laboratory in support of the VA’s Health Information Technology modernization. These detectors—Order Path Tracker, Trend Watcher, Pain Pointer, Performance Monitor, and Patient Record Flag Detector—leverage statistical and machine learning methods to monitor workflow disruptions, detect anomalies in care sequences and volumes, identify bottlenecks, and track system-level performance metrics across VistA and Millennium systems. All detectors have been integrated into the Health Data Analytics Platform (HDAP), with most having completed deployment and testing using live data from targeted stations in cardiology and oncology domains. This work enhances VA’s capacity for proactive system surveillance, promotes patient safety, and informs data-driven operational improvements across the EHR ecosystem.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Variability Analysis of FMVSS-121 Air Brake Systems: 60-mi/hr Service Brake System Performance Data for Truck Tractors

In support of the Federal Motor Carrier Safety Administration’s (FMCSA’s) ongoing interest in connected and automated commercial vehicles, this report summarizes analyses conducted to quantify variability in stopping distance tests conducted on commercial truck tractors. The data used were retrieved from tests performed under the controlled conditions specified for FMVSS-121 air brake system compliance testing. The report explores factors affecting the variability of the service brake stopping distance as defined by 49 CFR 571.121, S5.3.1 Stopping Distance—trucks and buses stopping distance. Variables examined in this analysis include brake type, weight, wheelbase, and tractor antilock braking system (ABS). This analysis uses existing test data collected between 2010 and 2019. Several of the examined parameters affected both tractor stopping distance and stopping distance variability. First, the average stopping distance for disc/disc brakes was shorter than either drum/drum or disc/drum brakes, but brake type did not have a statistically significant effect on stopping distance variability. Second, tractors with a gross vehicle weight rating (GVWR) of 45,000-50,000 lb had shorter stopping distances than any other examined weight category. Weight was found to have a significant effect on stopping distance variability; the 50,000-55,000-lb GVWR range had more variability than both the next lower (45,000-50,000 lb) and next higher (55,000-60,000 lb) range.

Lascurain, Mary Beth↗