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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 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Fracture toughness evaluation for thin-shell stainless steel weldment

Integrity of spent nuclear fuel (SNF) interim storage canisters are very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method of small specimen approach was used to estimate the canister weldment fracture toughness. Most SNTT weldment samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness $J_Q'$ for the baseline SS304 steel is at 283 kJ/m 2 . The estimated $J_Q'$ for the SS304/308 weld from HAZ and weld specimens are 126 kJ/m 2 and 148.9 kJ/m 2 , respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Analysis of Truck Platooning on Rural Highways

An analysis by the National Laboratory of the Rockies on truck platooning technology used on Ohio highways found that truck platooning operated for 40% of driving distances, showcasing its potential to enhance freight efficiency and safety. While energy savings were evident, further optimization of gap distances and operational consistency is needed to fully evaluate fuel savings benefits across diverse driving conditions. While more study is needed, advanced connected and automated vehicle technologies such as platooning demonstrate significant promise for transforming commercial vehicle efficiency and operations.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2021

Modular High Temperature Gas-Cooled Reactor: Safety Design Approach

Deployment Objectives Flexibly co-locate with new industry users of nuclear energy Steam and electric cogeneration applications Direct process heat with temperature ranges from 700°C to 950°C Enabling Requirements Meet regulatory dose limits at the Exclusion Area Boundary (EAB) 25 rem Total Effective Dose Equivalent (TEDE) for duration of the release from 10 CFR 50.34 (10 CFR 52.79) at Exclusion Area Boundary (EAB) for design basis accidents EAB is estimated approximately 400 meters from the modular HTGR plant (to support co-location with industrial facilities) Meet safety goals for cumulative individual risk for normal and off-normal operation Design goal: meet EPA Protective Action Guides (PAGs) at EAB 1 rem TEDE for sheltering Design basis and beyond design basis events are considered Realistically evaluated at the EAB Emergency planning and protection

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal Neutron Scattering Law Evaluations for Zirconium Carbide and Critical Mass Calculations [Slides]

Zirconium Carbide (ZrC) is a refractory material currently being investigated for high temperature reactor applications. ZrC was studied as a base of (U,Zr)C solid solution and UC 2 composite fuels for NERVA-type cores during the Rover program. Testing of the material was performed at Nuclear Furnace 1 (NF-1) located at Los Alamos. Thermal Scattering Law (TSL) evaluations generated with FLASSH are a new contribution to ENDF/B-VIII.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Informing Plant Asset Reliability and Availability Through AI-Driven Analysis of Operator Logs

The availability and reliability of nuclear power plant (NPP) structures, systems, and components (SSCs) are critical parameters for NPP safety. Tracking these parameters is necessary but costly and labor-intensive, requiring the collection and evaluation of SSC event data such as shutdowns, startups, and failures. To show how these events are needed for the parameters an example is given: one measure of reliability is based on the number of equipment failure events and the number of run hours (i.e., the time from a startup event to a shutdown event). Here, this work investigates using artificial intelligence (AI) to mine NPP operator log entry texts for SSC event data. Four AI approaches were explored for identifying these events, including natural language processing (NLP) methods, generative AI, generative AI combined with NLP, and topic modeling. A key challenge addressed with all four approaches is the brevity of operator log entries. Among these four a neural network–based NLP method was shown to be the most promising for this application, achieving F1 scores of 86.0% for shutdowns, 92.2% for startups, and 80.4% for failures on a subject-matter-expert-curated dataset from NPP operator logs, compared to a baseline of 66.6% for a random classifier. This shows that NLP methods can perform better than generative AI. Additionally, the NLP methods combined with generative AI were shown to perform better than generative AI alone. Generative AI was most successful at providing the background information for the NLP methods to use. This work demonstrates the potential to use AI to automate parameter collection from NPP operator log entries and other records.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Imaging of Calibrated Defects in Additively Manufactured Materials: Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components

Additive manufacturing (AM, or 3D printing) for commercial nuclear energy applications is an emerging method for cost-efficient manufacturing aimed at replacing aging nuclear reactor parts and reducing costs for new construction. Known material flaws in AM include low-density regions consisting of non-sintered powder, which have to be detected to ensure the safety of long-term performance nuclear reactor components. Currently, limited options are available for nondestructive evaluation (NDE), either during or post manufacturing. As a solution to NDE of AM, we are developing pulsed thermal imaging which is non-contact, one-sided, and scalable to arbitrary size and shape of the AM parts. Pulsed thermography (PT) system utilizes a high intensity flash lamp to rapidly heat surface of sample, and a high sensitivity fast-frame megapixel infrared (IR) camera to capture data of surface temperature variations. The acquired data cube consists of a stack of surface temperature images taken at different times. Information about material internal defects is extracted by analyzing the data cube. This report provides results of preliminary performance evaluation of pulsed thermal imaging capability in detection of imprinted flaws in AM metallic structures. The flaws were introduced into AM parts as imprinted hemispherical low density regions, consisting of trapped un-sintered metallic powder. Specimens for developed for this study consisted of AM stainless steel 316 and Inconel 718 plates. The diameters of imprinted defects varied from 1mm to 8mm, and their depths below the plate flat surface varied between 1mm and 6mm. Pulsed thermal tomography (PTT) processes the measured data cube to obtain 3D reconstructions of material effusivity using a unique inversion algorithm developed at Argonne. PTT has been previously used in imaging of similar size flat bottom hole (FBH) simulated defects in stainless steel 316 and Inconel 718 specimens. In the study involving AM specimens, PTT imaging results have shown that 1mm-diameter defects located 1mm and 2mm below the surface of specimens were detectable. Larger size defects were detectable at greater depth. We also explored an alternative approach to detection of material flaws in PT data cube, which is using neural learning-based approach to blind source separation. Detection of small material defects requires finding features in the data cube which have signal contrast levels approaching sensitivity limit of IR camera. In this study, an optimized Neural Learning based Blind Source Separation (NLBSS) algorithm, including Principal Component Analysis (PCA), and Independent Component Analysis (ICA) is demonstrated to automatically extract principal temporal and spatial features of thermography frames to enhance flaw detection. By using the NLBSS algorithm, material internal defects can be automatically detected. Furthermore, this processing approach compensates for experimental thermal imaging artifacts, such as noise and uneven heating. By merging artificial intelligence with phtotothermics, the NDE system detects internal calibrated defects of various sizes and depths in AM nuclear-grade metallic alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2021

181 Ta - fast neutron evaluation [Slides]

Reliable fast-neutron evaluation can be entirely encapsulated in the reaction model and related input. Ta181 is a particularly relevant case due to extraordinary coverage of various observables by differential experiments that offer much help and relatively little headache. Differences from the similar ENDF/B-VIII evaluation include: more advanced modeling (CC OMP, MSD, MSC, decay schemes), more careful selection of models and parameters, new experimental data, and isomers. Overall, improved agreement with differential data and hints of better performance in integral testing.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Updated ENDF-6 Format Approval Process [Slides]

The presentation begins by discussing the original ENDF/B-VIII.0 format proposal lifecycle and states that the GNDS Expert Group has developed a gitlab-friendly review process that addresses concerns raised by the format proposal lifecycle. As a result, a revised ENDF/B-VIII.0 format proposal lifecycle has been created. The presentation then discusses the necessary steps in the lifecycle that are as follows: create an issue in the issue tracker, make your changes directly to the manual, submit it for review and final approval. The presentation concludes by discussing two other things, including the transition CI/CD from buildbot system to built-in gitlab CI/CD and that the revised format review process document will follow, pending approval.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Proposed Generalized Header File for TSLs [Slides]

This presentation discusses several recent issues have been raised over the past few years regarding inconsistent definitions for the number of scattering atoms between incoherent elastic and inelastic (MT2 and MT4 files, respectively) in the same MF7 file and isotopic distribution of primary scattering atom. It also discusses how processing codes handle these potential issues in different ways. Additionally, it discusses how the preferred solution would be automation in order to reduce possibility of user error and a proposed solution in the form of a MF7/MT1 file is examined with preliminary examples. The presentation concludes with a question-and-answer discussion.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

FY22 CritView Status & Database Expansion [Slides]

This presentation covers Fiscal Year FY22 CritView Status & Database Expansion. This presentation Provide an update for progress and current status related to the CritView digital handbook for FY22/FY23. The lecture also covers a budget summary, FY22 highlights, CritView overview, and future work for FY23.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2025

Tantalum Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.02 cm Pitch)

The United States (US) Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories (SNL). The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water moderated and — reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM 096, LEU- COMP-THERM-097, LEU-COMP-THERM-101, LEU-COMP-THERM-102, and LEU-COMP-THERM-111. Applied interests in neutron-absorbing elements at epithermal and intermediate neutron energy ranges at DOE’s Hanford Site and Savannah River Site were identified through DOE Office of Environmental Management (EM) workshops and needs surveys performed in conjunction with the National Nuclear Security Administration/Nuclear Criticality Safety Program. In response, a team lead by Oak Ridge National Laboratory (ORNL) performed a study evaluating the use of 7uPCX for testing the epithermal cross sections of materials at SNL. This marked the inception of the set of experiments described here. The purpose of these experiments was to measure the effects of tantalum in nearly-critical systems. The tantalum was introduced into the fuel arrays as experiment rods within a central test region. The central test region was designed to target the epithermal neutron energy range by providing a dry cylindrical cavity constructed of aluminum and lined with 0.04 in (0.1016 cm) thick cadmium sheet. The central test region has an outer diameter of 3.75 in (9.525 cm) length of 31 in (78.74 cm) and can hold 85 tantalum rods nominally 0.25 in (0.635 cm) outside diameter and 31.25 in (79.375 cm) long. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 2016 fuel rod positions on a pitch of 0.4 in (1.016 cm) in a series of 21 hexagonal rings positioned around the central test region. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The eight cases in this experiment series were performed in 2024 in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility by varying the number of fuel rods in the critical assembly. The personnel involved in conducting these experiments included David Ames, Elijah Lutz, Beth Hanson, Patrick Ward, Jason Soares, and Augie Chapa from SNL, as well as Mathieu Dupont and B.J. Marshall from ORNL. Case 1 had no tantalum experiment rods. Case 2 had seven tantalum rods in the center of the central test region. Case 3 had 18 tantalum rods in the third hexagonal ring of the central test region. Case 4 had 19 tantalum rods in the center of the central test region. Case 5 had 30 tantalum rods in the outer most positions of the central test region. Case 6 had 37 tantalum rods in the center of the central test region. Case 7 had 61 tantalum rods in the center of the central test region. Case 8 had 85 tantalum rods filling all positions in the central test region. All eight critical experiments are judged to be acceptable as benchmark experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Comparative Evaluation of Control-Oriented Heavy Duty Vehicle Air Drag Coefficient Models

Heavy-duty vehicles (HDVs) are a significant source of fuel consumption and greenhouse gas emissions, prompting solutions such as HDV platooning to mitigate these negative impacts through air drag reduction. The intervehicle distance in an HDV platoon needs to be carefully selected, such that the platoon-level energy efficiency and safety considerations can be well balanced. Underlying this problem lies in accurately modeling the relationship between HDV air drag coefficient and intervehicle distance. Through comprehensive evaluation and comparison, we analyze five control-oriented HDV air drag coefficient models, including the polynomial model, rational polynomial model, rational model, semi-quadratic model, and ridge model. Leveraging Scipy Curve-Fit toolbox and our previously compiled air drag coefficient datasets, we optimally identify the parameters inside each model. The calibrated models are then thoroughly evaluated via five complementary metrics. The comparison results reveal that the semi-quadratic model has the highest overall performance, while the widely adopted rational model only exhibits suboptimal performance.

Best, Micah↗
DOE OSTI2025

Control Selection for the Neutralization Tank in the Aqueous Recovery System at the Savannah River Plutonium Processing Facility

The Aqueous Recovery System (ARS) at the Savannah River Plutonium Processing Facility (SRPPF) recovers and purifies plutonium (Pu) using aqueous chemistry techniques. The neutralization tanks in the ARS collect the waste streams, containing various impurities and greatly reduced concentrations of Pu, that are produced during aqueous processing. Criticality safety is primarily achieved in the ARS by using geometrically favorable process vessels. However, due to the geometry of neutralization tanks, each tank is administratively limited to no more than 450 g Pu, thus representing the transition from geometry to mass control. To ensure that this mass limit is not exceeded, an administrative control requires taking two samples of any solution to be sent to a neutralization tank. Normal operations are expected to result in less than 50 g Pu in a neutralization tank filled to its capacity of 250 L. Process upsets may cause an inadvertent transfer of up to 1000 g Pu to a neutralization tank, producing a system that is potentially not subcritical for all possible configurations of the tank. To ensure that a neutralization tank remains safety subcritical, passive engineered controls (e.g.changing tank geometry or adding fixed poisons), active engineered controls (e.g. interlocks), and administrative controls (e.g. soluble poisons and valve isolation) were considered and their viability evaluated. Ultimately, the team chose an administrative, dual-valve isolation strategy. This paper will thoroughly discuss the various control strategy options and why many of the options were not feasible for maintaining criticality safety.

Dressman, Phillip M. [Savannah River Nuclear Solut↗
DOE OSTI2020

Sensitivity Studies, Gap Analysis, and Benchmark Experiment Optimization for Reactor Physics and Criticality Safety Applications

Many new reactor designs, such as advanced reactors and micro reactors, have materials that lack nuclear data validation. This is also true for many other applications in criticality safety and global security. Both differential and integral experiments are needed to validate cross-section data. Without this, a user cannot have confidence in the predicted results of a radiation-transport code. This work describes an approach called ARCHIMEDES (Application Relevant Critical/Subcritical HEU/Pu-based Integral Measurements for Enhancing Data and Evaluating Sensitivities) to design new criticality experiments that have similar k eff cross-section sensitivities to an application of interest. This process involves simulations to generate cross-section sensitivities to a parameter of interest (such as k eff ), a gap analysis to determine which existing benchmarks are most similar to the application, and an experiment optimization. Recently, there has been a great deal of interest in the reactor physics community on advanced reactors, micro reactors, and accelerator driven systems (ADS). This work will apply the described method to specific examples in this area. The focus of this work will be on the sensitivity study and gap analysis, while future work will include experiment design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Non-dimensional performance and safety parameters for heat pipes

The use of heat pipes in safety-critical systems such as nuclear microreactors dictates the development of generalized, practical, scalable performance and safety parameters. Traditional dimensional metrics, while informative, lack the universality required for comparative analysis across varying designs and operating regimes. Here, this work introduces a comprehensive set of non-dimensional parameters to characterize heat pipe performance and safety, including capillary performance, effective thermal conductivity, response time, exergetic efficiency, allowable temperature gradients, allowable rate of temperature change, priming coefficients, and factor of safety. A reference heat pipe design representative of microreactor applications was analyzed via the developed parameters using both traditional analytical models and Sockeye simulations under transient and steady-state conditions. Sodium, potassium, and water were evaluated as working fluids to demonstrate the applicability of the framework across a broad temperature range. The proposed non-dimensional parameters effectively captured key thermal-hydraulic behaviors and safety concerns, as was demonstrated via Sockeye simulations. This framework supports the development of design optimization strategies, operational protocols, and safety assurance practices for advanced reactor systems and other high-reliability applications.

42 - ENGINEERING↗
DOE OSTI2025

Control Selection for the Neutralization Tank in the Aqueous Recovery System at the Savannah River Plutonium Processing Facility

The Aqueous Recovery System (ARS) at the Savannah River Plutonium Processing Facility (SRPPF) recovers and purifies plutonium (Pu) using aqueous chemistry techniques. The neutralization tanks in the ARS collect waste streams with various impurities and greatly reduced concentrations of Pu than are produced during aqueous processing. Criticality safety is primarily achieved in the ARS by using geometrically favorable process vessels. However, due to the geometry of neutralization tanks, each tank is administratively limited to no more than 450 g Pu, thus representing the transition from geometry to mass control. To ensure that this mass limit is not exceeded, an administrative control requires taking two samples of any solution to be sent to a neutralization tank. Normal operations are expected to result in less than 50 g Pu in a neutralization tank filled to its capacity of 250 L. Process upsets may cause an inadvertent transfer of up to 1000 g Pu to a neutralization tank, producing a system that is potentially not subcritical for all possible configurations of the tank. To ensure that a neutralization tank remains safety subcritical, passive engineered controls (e.g. changing tank geometry or adding fixed poisons), active engineered controls (e.g. interlocks), and administrative controls (e.g. soluble poisons and valve isolation) were considered and their viability evaluated. Ultimately, the team chose an administrative, dual-valve isolation strategy. This paper will thoroughly discuss the various control strategy options and why many of the options were not feasible for maintaining criticality safety.

Dressman, Phillip M. [Savannah River Nuclear Solut↗