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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

PermitTEC v0.1: Standardized Metadata Corpus of NEPA Litigation Documents

The National Environmental Policy Act of 1969, as amended (NEPA), mandates that federal agencies assess and document potential environmental impacts before deciding on proposed actions. While significant progress has been made in cataloging and standardizing NEPA documents themselves, the legal challenges that frequently arise from these decisions remain poorly cataloged and largely inaccessible for systematic analysis. Litigation challenging NEPA compliance can substantially delay project timelines, reshape agency decision-making, and establish precedents that influence future environmental reviews — yet no standardized, machine-readable corpus exists that links litigation records to the NEPA projects they contest.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

DeepPDF: A Deep Learning Approach to Extracting Text from PDFs

Scientific publications contain a plethora of important information, not only for researchers but also for their managers and institutions. Many researchers try to collect and extract this information in large enough quantities that it requires machine automation, but because publications were historically intended for print and not machine consumption, the digital document formats used today (primarily PDF) have created many hurdles for text extraction. Primarily, tools have relied on trying to convert PDF's to plain text documents for machine processing by reverse engineering the PDF standard. A complex process because once a PDF is created it is more closely related to an image file than a document markup language. In this paper we explore the feasibility of treating these PDF documents as images as opposed to a proprietary markup language. We believe that by using deep learning and image analysis we can create more accurate PDF to text extraction tools than those that currently exist. \\ \newline \Keywords{deep learning, text extraction, information extraction, PDF extraction, scholarly publications.

Stahl, Christopher↗
DOE OSTI2023

Topics in Optical Design, Manufacturing, and Analysis

Presentation to be given at the LANL Detectors and Analysis Workshop December 5-6, 2023. This STIP is a resubmission of STIP 35977839 (an incorrect version of the document was sent through the STIP process previously).

42 ENGINEERING↗
DOE OSTI2017

Consideration for Implementing the Safety Regulation Plans on Permanent Shut-own NPP during Transition Period in Korea

Recently the Korean regulatory body has been reviewing the change of permission on permanent shutdown Kori NPP (Nuclear Power Plant) unit 1, since the licensee (KHNP, Korea Hydro and Nuclear Power Co. Ltd) decided the permanent shut-down in June 2015 and applied the amendment of its licensing basis documents in June 2016. Although our institute with technical authority, the KINS (Korea Institute of Nuclear Safety), had performed the peer reviews for the construction permits of 3 NPPs and operating licenses of 25 NPPs, the safety control of permanent shut-down NPP and the transition activities including deactivation and decommissioning preparation are still in early stages and challenging fields. Currently the operation/only-possession system under defueled condition were defined through the peer review of fully revised documents and the major issues which concerned with the rearrangement of safety class or function based on new accident analysis results were produced. And also various transition activities were drawn as reflecting the decommissioning strategy on Kori unit 1. In this study, we described the status and problem of safety regulation based on our subordinate statute for permanent shut-down and transition activities. And then the detail implementation plans or the alternatives will be provided and discussed in order to solve the lack of regulatory experience and guide.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Thermal Model of Oxide Growth for Full Size Fuel Plate Experiments in the Advanced Test Reactor

The full-size plate number one (FSP-1) irradiation test designed for irradiation testing of plate type fuel is planned to be irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). This FSP-1 experiment is a non-instrumented drop-in test where aluminum-clad fuel plates are cooled directly by the ATR Primary Coolant System (PCS) water. This experiment is one of a series of experiments being irradiated at ATR for the United States High Performance Research Reactor (USHPRR) program. This fueled experiment contains aluminum-clad fuel full-size plates consisting of monolithic U-10Mo. Previous thermal analyses have been documented concerning oxide growth [1] and thermal safety margins [2] for similar U-10Mo mini plate experiments. This analysis is part of the thermal safety analysis of the experiment to be irradiated in ATR. Departure from nucleate boiling ratio (DNBR) and flow instability ratio (FIR) have been calculated for the flow coastdown condition 2 and the reactivity insertion accident condition-2 (RIA2) transient. The purpose of this paper is to investigate the thermal margin of the fuel meat during a RIA2 transient when oxide has accumulated on the plate surfaces acting as an insulator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Selection of Nuclear Fuel for TREAT: UO 2 vs U 3 O 8

The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight). The TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore, to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for the fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO 2 vs U 3 O 8 . In this report, the results are documented pertaining to the choice mentioned above (UO 2 vs U 3 O 8 ). The conclusion in favor of using UO 2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and current literature were reviewed. This analysis was supplemented by the construction and examination of the U-O and U-O-C phase diagrams at pressure close to negligent, thereby mimicking the conditions in which nuclear fuel is supposed to function inside the zirconium-based cladding in the reactor. Finally, our conclusion in favor of the UO 2 down selection was summarized and explained in the last Section of this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Code Coverage Status of the ARC code GAMSRC

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. GAMSOR serves as procedure to obtain the neutron and gamma power distribution information in the ARC code system. GAMSOR is a specially modified version of DIF3D (dif3d.x becomes dif3d_gamsor.x). After some work, it was determined that carrying out software on GAMSOR was impractical and would not fit well with commercial grade dedication. GAMSRC was written in the last decade to replace GAMSOR such that the user can rely soley upon the verified DIF3D (dif3d.x) code and GAMSRC to obtain the neutron and gamma power distribution information. GAMSRC is also fully verified and ready for commercial grade dedication. In the coming years, GAMSOR will be deprecated and GAMSRC will fully take over in the ARC code system. This document identifies the set of test problems used to assess the code coverage for GAMSRC. The goal is to document what parts of the existing GAMSRC code are touched by the set of test problems and which are not. The code coverage analysis of GAMSRC was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of GAMSRC. The code coverage tables are given for each submodule of GAMSRC. Because GAMSRC links to modules in DIF3D, some details on coverage changes to the DIF3D lined files is provided. As will be seen, most of the uncovered parts/files can be ignored because they are either for error message and debugging output or obviously not needed by GAMSRC today.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

SNL Building - High Speed Computer: Program Requirements Document, Rev 1

This PRD is developed for multiple delivery options which have yet to be determined: The intent for the options is to allow for a risk analysis to be performed and determine the most appropriate location. Each location has benefits where others may not. These options will be vetted in a workshop December 7 and 8, 2016. Once an option is selected the requirements pertaining to options that are not selected will be removed from this PRD.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

User Manual for Whisper-1.1

Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation [1-3]. It uses the sensitivity profile data for an application as computed by MCNP6 [4-6] along with covariance files [7,8] for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014 [3]. During 2015- 2016, Whisper was updated to version 1.1 and is to be included with the upcoming release of MCNP6.2. This document describes the user input and options for running whisper-1.1, including 2 perl utility scripts that simplify ordinary NCS work, whisper_mcnp.pl and whisper_usl.pl. For many detailed references on the theory, applications, nuclear data & covariances, SQA, verification-validation, adjointbased methods for sensitivity-uncertainty analysis, and more – see the Whisper – NCS Validation section of the MCNP Reference Collection at mcnp.lanl.gov. There are currently over 50 Whisper reference documents available.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Evaluation of Engineered Barrier Systems in the Disposition of Spent Nuclear Fuel

This document is a summary of the R&D activities associated with the Engineered Barrier Systems Work Package. Multiple facets of Engineered Barrier Systems (EBS) research were examined in the course of FY18 activities. This report is focused on delivering an update on the status and progress of modelling tools and experimental methods, both of which are essential to understanding and predicting long-term repository performance as part of the safety case. Specifically, the work described herein aims to improve understanding of EBS component evolution and interactions. Utlimately, the EBS Work Package is working towards producing process models for distinct processes that can either be incorporated into performance assessment (PA), or provide critical information for implementing better constraints on barrier performance The main objective of this work is that the models being developed and refined will either be implemented directly into the Generic Disposal System Analysis platform (GDSA), or can otherwise be indirectly linked to the performance assessment by providing improved bounding conditions. In either the case, the expectation is that validated modelling tools will be developed that provide critical input to the safety case. This report covers a range of topics — modelling topics include: thermal-hydrologic-mechanical-chemical coupling (THMC) in buffer materials, comparisons of modelling approaches to optimize computational efficiency, thermal analysis for EBS/repository design, benchmarking of thermal analysis tools, and a preliminary study of buffer re-saturation processes. Experimental work reported, includes: chemical evolution and sorption behavior of clay-based buffer materials and high-pressure, high temperature studies of EBS material interactions. The work leverages international collaborations to ensure that the DOE program is active and abreast of the latest advances in nuclear waste disposal. This includes participation in the HotBENT Field Test, aimed at understanding near-field effects on EBS materials at temperatures above 100 °C, and the analysis of data and characterization of samples from the FEBEX Field Test. Both the FEBEX and HotBENT Field Tests utilize/utilized the Grimsel Test Site in Switzerland, which is situated in a granite host rock. These tests offer the opportunity to understand near field evolution of bentonite buffer at in situ conditions for either a relatively long timescale (18 years for FEBEX) or temperature above 100 °C (HotBENT). Overall, this report provides in depth descriptions of tools and capabilities to investigate nearfield performance of EBS materials (esp. bentonite buffer), as well as tools for drift-scale thermal and thermal-hydrologic analysis critical to EBS and repository design. For a more detailed description of work contained herein, please see Section 10 ("Conclusions") of this document.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Biorefinery upgrading of herbaceous biomass to renewable hydrocarbon fuels, Part 2: Air pollutant emissions and permitting implications

The development of advanced biofuel production facilities is still at a nascent stage, and the biorefineries could face challenges in obtaining air permits required for their construction and operation because they are novel emission sources. To fill gaps in knowledge regarding potential emissions, we perform a detailed federal regulatory analysis and estimate air pollutant emissions for an advanced biorefinery that produces renewable diesel blendstock (RDB) from lignocellulosic biomass via aerobic respiration documented in Part 1 of the paper. We evaluate 12 design permutations that include two feedstocks, a uniform format blend (UFB) and corn stover; three biorefinery scales, 2,000, 5,200, and 9,100 dry metric tons per day (dmtd) of lignocellulosic biomass feed; and two lignin uses, as either a boiler fuel or for pellet production. We also evaluate 6 additional design permutations by incorporating non-emitting renewable power, which could reduce up to 63% carbon monoxide (CO) and nitrogen oxides (NO x ) each, 21% volatile organic compounds (VOC), and 43% hazardous air pollutants (HAP) as opposed to on-site power production, for biorefineries that produce pellets, using either UFB or corn stover. Our results indicate that all 18 design permutations would be classified as a major source under the Clean Air Act's New Source Review program without additional emission controls. Compared to using lignin as a boiler fuel, diverting lignin for pellet production reduces the emissions of CO up to 88%, NOx up to 73%, sulfur dioxide (SO 2 ) up to 99%, VOCs up to 72%, and HAPs up to 66%. Additionally, we explore control options that could further reduce emissions and assess whether reductions are enough to achieve minor source classification. Insights from our analysis can help biorefinery developers and regulators develop permitting strategies to mitigate investment risks.

09 BIOMASS FUELS↗
DOE OSTI2019

A combined static-dynamic single-dose imaging protocol to compare quantitative dynamic SPECT with static conventional SPECT

Background: SPECT myocardial perfusion imaging (MPI) is a clinical mainstay that is typically performed with static imaging protocols and visually or semi-quantitatively assessed for perfusion defects based upon the relative intensity of myocardial regions. Dynamic cardiac SPECT presents a new imaging technique based on time-varying information of radiotracer distribution, which permits the evaluation of regional myocardial blood flow (MBF) and coronary flow reserve (CFR). In this work, a preliminary feasibility study was conducted in a small patient sample designed to implement a unique combined static-dynamic single-dose one-day visit imaging protocol to compare quantitative dynamic SPECT with static conventional SPECT for improving the diagnosis of coronary artery disease (CAD). Methods: Fifteen patients (11 males, four females, mean age 71 ± 9 years) were enrolled for a combined dynamic and static SPECT (Infinia Hawkeye 4, GE Healthcare) imaging protocol with a single dose of {sup 99m}Tc-tetrofosmin administered at rest and a single dose administered at stress in a one-day visit. Out of 15 patients, eleven had selective coronary angiography (SCA), 8 within 6 months and the rest within 24 months of SPECT imaging, without intervening symptoms or interventions. The extent and severity of perfusion defects in each myocardial region was graded visually. Dynamically acquired data were also used to estimate the MBF and CFR. Both visually graded images and estimated CFR were tested against SCA as a reference to evaluate the validity of the methods. Results: Overall, conventional static SPECT was normal in ten patients and abnormal in five patients, dynamic SPECT was normal in 12 patients and abnormal in three patients, and CFR from dynamic SPECT was normal in nine patients and abnormal in six patients. Among those 11 patients with SCA, conventional SPECT was normal in 5, 3 with documented CAD on SCA with an overall accuracy of 64%, sensitivity of 40% and specificity of 83%. Dynamic SPECT image analysis also produced a similar accuracy, sensitivity, and specificity. CFR was normal in 6, each with CAD on SCA with an overall accuracy of 91%, sensitivity of 80%, and specificity of 100%. The mean CFR was significantly lower for SCA detected abnormal than for normal patients (3.86±1.06 vs 1.94±0. 0.67, P < 0.001). Conclusions: The visually assessed image findings in static and dynamic SPECT are subjective, and may not reflect direct physiologic measures of coronary lesion based on SCA. The CFR measured with dynamic SPECT is fully objective, with better sensitivity and specificity, available only with the data generated from the dynamic SPECT method.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2025

A Performant, Scalable Processing Pipeline for High‐Quality and FAIR Environmental Sensor Data

High-resolution environmental monitoring is necessary to record, understand, and predict biogeochemical and ecological changes particularly in coastal systems but brings significant challenges in processing and making rapidly available the resulting data. The COMPASS-FME project established a network of coastal observational sites across the Chesapeake Bay and western Lake Erie regions extensively instrumented with soil, vegetation, and weather sensors logging data every 15 min. Our data processing framework, written in R and completely open source, prioritizes rapid model-experiment iteration and makes biogeochemical data rapidly available for quality assurance/quality control, analysis, and model ingestion. This pipeline is distinguished by a standardized and modular approach to data curation, extensive metadata and documentation, and its high performance. These attributes combine to make biogeochemical data rapidly accessible across COMPASS-FME and the broader community. Flexible, powerful, and reproducible approaches to handling high-volume environmental data are crucial for accelerating biogeosciences research.

Pennington, Stephanie C. [Pacific Northwest Nation↗
DOE OSTI2025

FAST Irradiations, Postirradiation Examinations, and Modeling of U-Mo for Light Water Reactor Applications

Many next generation light water reactor (LWR) concepts, such as mobile small modular reactors, are seeking to use smaller core dimensions than conventional reactor types. Smaller reactor cores require an increase in fissile material to maintain reactivity. For non-proliferation purposes, enrichment increases are limited to less than 20% (high assay low enriched uranium, [HALEU]) and so higher uranium density fuels than UO 2 must be considered. To this end, uranium-molybdenum alloys were tested using the Fission Accelerated Steady-state Test (FAST) approach. The experiment test matrix is focused on identifying the temperature transition between low swelling and high fission gas retention to break away swelling and low fission gas retention. This paper documents the results of irradiation tests and post-irradiation examinations (PIE) including neutron radiography, rodlet profilometry, fission gas collection analysis, and optical metallography. The results of these tests showed that unconstrained U-Mo fuels (solid, Na-bonded rodlets) have a swelling threshold between 400-450°C with minimal fission gas release below this point. Higher temperature solid fuel showed microstructural zoning with small pore networks while lower temperature solid fuels have a uniform microstructure with large pore networks. U-Annular Mo fuels where swelling had some self-constraint imposed upon it, were shown to have much reduced swelling compared to their solid counterparts as well as very low fission gas release for irradiation temperatures up to 500°C. These initial results show that the use of U-Mo in constrained fuel geometries could be used as a high uranium density HALEU fuel for LWRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Evaluation of Repurposing Archetypal Preventive Radiological/Nuclear Detectors to the Consequence Management Mission

In the first hours or days after an unplanned release of radioactive material to the environment, the radiation detection instruments most widely available to local first responders may be those currently fielded for interdiction missions. This study investigated how such preventative radiological / nuclear detection instruments could perform if repurposed to consequence management missions. A representative sample of three archetypes (body-worn, human carried, and other/large detection volume) encompassed six categories: personal radiation detector, extended range personal radiation detector, personal emergency radiation detector, radio-isotope identification device, human portable detector/backpack, and vehicle mounted large detection volume. Overall 19 models of equipment were included in the study. Laboratory evaluations were designed to assess the capabilities of the instruments in four consequence management missions: exposure rate, integrated exposure, radiation survey, and contamination screening. As applicable, the evaluations included measurement of exposure rate, integrated exposure, overrange response, and angular response. The results were compared to benchmarks from the American National Standards Institute ANSI N42.49A. The performance of the instruments for initial screening for contamination was assessed by an automated radioactive source moving past the detectors at various speeds and distances. The results demonstrate that if the equipment is used in accordance with the mission analysis and categories, and within the original equipment manufacturer specifications, it is possible to achieve sufficient accuracy to estimate and document doses to responders, plan entries into contaminated areas, detect contamination, and protect the public, until such time as outside resources arrive with sufficient numbers of standard health physics instruments and personnel dosimetry to replace the preventative radiological / nuclear detection instruments. Finally, this evaluation campaign was conducted to complement the National Council on Radiation Protection and Measurements Report-179 on Guidance for Emergency Response Dosimetry.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

PCF File Format.

PCF files are binary files designed to contain gamma spectra and neutron count rates from radiation sensors. It is the native format for the GAmma Detector Response and Analysis Software (GADRAS) package [1]. It can contain multiple spectra and information about each spectrum such as energy calibration. This document outlines the format of the file that would allow one to write a computer program to parse and write such files.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

FY2017 Report on NISC Measurements and Detector Simulations

FY17 work focused on automation, both of the measurement analysis and comparison of simulations. The experimental apparatus was relocated and weeks of continuous measurements of the spontaneous fission source 252 Cf was performed. Programs were developed to automate the conversion of measurements into ROOT data framework files with a simple terminal input. The complete analysis of the measurement (which includes energy calibration and the identification of correlated counts) can now be completed with a documented process which involves one simple execution line as well. Finally, the hurdles of slow MCNP simulations resulting in low simulation statistics have been overcome with the generation of multi-run suites which make use of the highperformance computing resources at LANL. Preliminary comparisons of measurements and simulations have been performed and will be the focus of FY18 work.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Bison Robustness and Performance Improvements

In last year’s Bison robustness report, the authors demonstrated that improving the Jacobian matrix can drastically increase solution convergence rates. The development of an automatic differentiation (AD) capability in the MOOSE framework has created the possibility of having exact Jacobian matrices for all of our material models. A large focus in this year’s report is the application of AD to MOOSE physics modules, such as the tensor mechanics module, and the application to Bison fuel and cladding material models. A second main focus in this year’s report is the investigation of a set of algorithmic solver changes to improve the robustness and performance of Bison LWR simulations. Beyond providing increased solver robustness, the ability to solve strongly coupled multiphysics problems, and solving large deformation problems with ease, these automatically generated perfect Jacobians also drastically reduce development time for new models. In models converted to AD, Jacobian computation made up the majority of the code base. With this code removed, both readability and maintainability of the MOOSE/Bison code base are improved. The work documented here significantly improved the robustness of the Bison LWR assessment case suite. Through algorithmic changes in solver infrastructure, such as the application of predictors and improvements in timestepping, tangible improvements in performance were observed. These improvements are documented using a new standard set of metrics that are collected in all assessment cases, and a suite of statistical analysis tools developed for this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗