Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Nuclear Test”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Data-driven reevaluation of f t values in superallowed β decays

We present a comprehensive reevaluation of the ft values in superallowed nuclear β decays crucial for the precise determination of V ud and low-energy tests of the electroweak standard model. It consists of the first, fully data-driven analysis of the nuclear β decay form factor, that utilizes isospin relations to connect the nuclear charged weak distribution to the measurable charge distributions. This prescription supersedes previous shell-model estimations, and allows for a rigorous quantification of theory uncertainties in f which is absent in the existing literature. Furthermore, our new evaluation shows an overall downward shift of the central values of f at the level of 0.01%.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Annual Performance Testing of Tracer Gas and Tracer Aerosol Detectors for use in Radionuclide NESHAP Compliance Testing

Los Alamos National Laboratory’s (LANL’s) compliance with Radionuclide NESHAP 1 regulations is managed by the Radionuclide Air Emissions Management (RAEM) team, part of LANL’s Compliance Programs group (EPC-CP). One area of the Radionuclide NESHAP addresses requirements for siting a stack sample system. This memo documents the most recent performance tests of the sulfur hexafluoride detectors for use in stack gas mixing tests, as well as the functional trending test on a pair of optical particle counters, comparing results from a non-calibrated instrument to a calibrated instrument.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Reliability of the {sup 123}I-mIBG heart/mediastinum ratio: Results of a multicenter test–retest reproducibility study

A quantitative measurement, the Heart-to-Mediastinum (H/M) ratio of counts derived from a planar acquisition approximately 4 hours after injection of {sup 123}I-mIBG, is a strong predictor of outcomes in patients with stable class II-III heart failure and LVEF ≤ 35%. This study assessed the test–retest reproducibility of the H/M ratio in such patients. 47 subjects with class II-III systolic heart failure and LVEF ≤ 35% were tested at two time intervals separated by 5 to 14 days. Subjects were imaged twice on the same camera using the same radionuclide dose. Images were sent to a core analysis lab, where three nuclear technologists independently determined the H/M ratios. The primary endpoint was test–retest H/M ratio reproducibility calculated as the absolute difference in mean value determined by the three readers. Mean subject age was 65 ± 12 years, 85% were male, and mean BMI was 29 ± 6 kg/m{sup 2}. Mean injected activity was 10.18 ± 0.43 mCi for first dose and 10.09 ± 0.52 mCi for the second dose. The mean and SD values for first and repeat studies were almost identical: the 95% confidence interval of the mean test–retest difference was 0.055 to 0.076. Bland–Altman plots showed no systematic effect of the H/M ratio on the magnitude of the difference between replicate measurements. Inter-reader measurements were nearly identical. There were no serious adverse events despite exposure to {sup 123}I-mIBG on 2 occasions in a short time period. The Heart-to-Mediastinum ratio of {sup 123}I-mIBG is a consistent and highly reproducible measurement in stable Class II to III heart failure patients.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2024

Using Generative AI to implement the discrepancy checker for a Nearly Autonomous Management and Control System for Advanced Reactors

Developments related to generative artificial intelligence (AI) have brought a major breakthrough in AI. These developments are rapidly accelerating developments in different science and engineering applications. Nearly Autonomous Management and Control (NAMAC) system provides recommendations to the operator for maintaining the safety and performance of the reactor. The discrepancy checker (DC) is an important component of the NAMAC) system, whose goal is to determine if the plant is moving towards the expected system state after the control actions are injected. In this work, we explore generative AI methods, particularly, a generative pretrained transformer (GPT) for implementing the DC function in NAMAC. The GPT-based DC aims to alert the operator in situations outside NAMAC’s scope and act as a chatbot the operator can use to retrieve relevant information. This study involves two versions of GPT developed by OpenAI: GPT-3.5 and GPT-4. These GPTs are trained on huge amounts of undisclosed general domain datasets. We explored two methods to adapt GPTs for DC implementation in NAMAC: fine-tuning and retrieval augmented generation. A small knowledge base (information file) that encompasses rules for DC implementation and some general information related to NAMAC has been created to support DC implementation using GPT. In this work, the GPT-based DC implementations have been tested for their reasoning abilities, comprehension, information retrieval, and extraction abilities. It should be noted that this paper only presents a preliminary study to test the feasibility of DC implementation using generative AI technology. Given the potential risks and severe consequences associated with nuclear reactor applications, combined with the black-box nature of AI, extensive offline and online testing and reliability analyses of GPT-based DCs are needed for further developing such capabilities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Instrumentation for the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) Project

Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), the French Atomic and Alternative Energies Commission (CEA), the Joint Research Center (JRC), the Nuclear Research and Consultancy Group (NRG), and the Research Center Rez (CVR), started a Joint Experimental Program (JEEP) project that operates within the Nuclear Energy Agency’s Framework for Irradiation Experiments (FIDES II) program in order to develop capabilities for the in-core real-time mechanical testing of structural materials. This effort will focus on designing a shared capsule capable of housing a variety of in-core mechanical testing instrumentation allowing enhanced experiments for the material science community. The outcome of the project would be high-priority, stress relaxation data for stainless-steel-based materials provided by EPRI and CEA. Stress relaxation is a major phenomenon that contributes to material degradation in nuclear reactor components. Currently, nuclear material stress relaxation is assessed both before and after irradiation, using complex and costly post-irradiation examination (PIE) activities. In-situ data would support the development of precision modeling and simulation of this degradation phenomena and would provide validation and benchmarking for existing models using the PIE data. As part of the U.S. Department of Energy Advanced Sensor and Instrumentation (ASI) program, INL has fabricated and tested out-of-core mechanical test instrumentation. This instrumentation was designed for easy adaptation to the irradiation capsule proposed under this JEEP, and can be deployed to measure real-time stress relaxation under pressurized-water reactor (PWR) conditions. This initial effort will partially serve to replace the testing capabilities lost because of shutting down the Halden Boiling Water Reactor (HBWR). The project would provide these capabilities to the international community via a shared capsule design that is easily adaptable to additional material test reactors. The design features will incorporate expansion to PWR and non-light-water reactor (LWR) environments that will be developed in future work. The capsule and instrumentation will be demonstrated in the Massachusetts Institute of Technology Reactor (MITR) for phase I and the Petten High Flux Reactor (HFR) for phase II irradiations to deliver real-time stress relaxation data on high priority stainless steel structural materials.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Metastable nuclear isomers as dark matter accelerators

Inelastic dark matter and strongly interacting dark matter are poorly constrained by direct detection experiments since they both require the scattering event to deliver energy from the nucleus into the dark matter in order to have observable effects. We propose to test these scenarios by searching for the collisional deexcitation of metastable nuclear isomers by the dark matter particles. The longevity of these isomers is related to a strong suppression of γ- and β-transitions, typically inhibited by a large difference in the angular momentum for the nuclear transition. The collisional deexcitation by dark matter is possible since heavy dark matter particles can have a momentum exchange with the nucleus comparable to the inverse nuclear size, hence lifting tremendous angular momentum suppression of the nuclear transition. This deexcitation can be observed either by searching for the direct effects of the decaying isomer, or through the rescattering or decay of excited dark matter states in a nearby conventional dark matter detector setup. Existing nuclear isomer sources such as naturally occurring 180m Ta, 137m Ba produced in decaying Cesium in nuclear waste, 177m Lu from medical waste, and 178m Hf from the Department of Energy storage can be combined with current dark matter detector technology to search for this class of dark matter.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

TREAT LEU Conversion Analyses Performed 2016-2018

TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. After more than 20 years of non-operational standby, TREAT has been restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT from its current highly enriched uranium fuel to a low enriched uranium fuel is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization Reactor Conversion Program. This report documents additional analyses performed over the timeframe of 2016 to early 2018 to evaluate the LEU Conceptual Design for the TREAT reactor, following the initial design studies documented in the 2016 Conceptual Design Report. The analyses documented in this report include neutronic and thermal hydraulic sensitivity studies to evaluate the effect of key design parameters on predicted core behaviors, as well as a simple investigation into the effect of fuel thermal expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

The Evaluation of Machine Learning Techniques for Isotope Identification Contextualized by Training and Testing Spectral Similarity

Precise gamma-ray spectral analysis is crucial in high-stakes applications, such as nuclear security. Research efforts toward implementing machine learning (ML) approaches for accurate analysis are limited by the resemblance of the training data to the testing scenarios. The underlying spectral shape of synthetic data may not perfectly reflect measured configurations, and measurement campaigns may be limited by resource constraints. Consequently, ML algorithms for isotope identification must maintain accurate classification performance under domain shifts between the training and testing data. To this end, four different classifiers (Ridge, Random Forest, Extreme Gradient Boosting, and Multilayer Perceptron) were trained on the same dataset and evaluated on twelve other datasets with varying standoff distances, shielding, and background configurations. A tailored statistical approach was introduced to quantify the similarity between the training and testing configurations, which was then related to the predictive performance. Wilcoxon signed-rank tests revealed that the OVR-wrapped XGB significantly outperformed the other algorithms, with confidence levels of 99.0% or above for the 133Ba, 60Co, 137Cs, and 152Eu sources. The findings from this work are significant as they outline techniques to promote the development of robust ML-based approaches for isotope identification.

domain adaptation↗
DOE OSTI2022

Operation Plumbbob

On June 24, 1957, a thirty-seven kiloton nuclear device, codenamed Priscilla, exploded 700 feet over the Nevada Test Site’s Frenchman dry lake. Below, on the lakebed, was a bank vault designed by the Mosler Safe Company for one of several Federal Civil Defense Administration experiments. The vault was constructed “to evaluate the effectiveness of a reinforced-concrete vault and steel vault door in providing protection against the effects of a nuclear detonation.” Blast damage to the exterior was largely confined to the side walls, whose concrete was stripped off; its steel reinforcing rods bent and twisted. The vault door was essentially undamaged and was opened with minimal difficulty. The vault contents - sandbags, movie camera, and thermometers - were only slightly disturbed. The camera operated as planned, running 40 ft of film. The film, however, was completely fogged. Two clinical type thermometers registered 88°F, and a 24-hr stylus American temperature recorder registered a constant 88°F. Priscilla was the fifth of twenty-nine tests conducted in 1957 as part of Operation Plumbbob, whose stated purpose was to conduct tests “aimed to attain new knowledge important to the defense of the United States and the Free World.” Individual events included “proof tests of air defense and antisubmarine warheads to be stockpiled, development tests of bomb components and mockups, and design tests of smaller and lighter warheads.” One test, John, was that of an air-to-air missile detonated 18,500 feet over the test site.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

thornado-hydro: A Discontinuous Galerkin Method for Supernova Hydrodynamics with Nuclear Equations of State

This paper describes algorithms for non-relativistic hydrodynamics in the toolkit for high-order neutrino radiation hydrodynamics (thornado), which is being developed for multiphysics simulations of core-collapse supernovae (CCSNe) and related problems with Runge–Kutta discontinuous Galerkin (RKDG) methods. More specifically, thornado employs a spectral type nodal collocation approximation, and we have extended limiters — a slope limiter to prevent non-physical oscillations and a bound-enforcing limiter to prevent non-physical states — from the standard RKDG framework to be able to accommodate a tabulated nuclear equation of state (EoS). To demonstrate the efficacy of the algorithms with a nuclear EoS, we first present numerical results from basic test problems in idealized settings in one and two spatial dimensions, employing Cartesian, spherical-polar, and cylindrical coordinates. Then, we apply the RKDG method to the problem of adiabatic collapse, shock formation, and shock propagation in spherical symmetry, initiated with a 15 M ⊙ progenitor. Herein, we find that the extended limiters improve the fidelity and robustness of the RKDG method in idealized settings. The bound-enforcing limiter improves robustness of the RKDG method in the adiabatic collapse application, while we find that slope limiting in characteristic fields is vulnerable to structures in the EoS — more specifically, in the phase transition from nuclei and nucleons to bulk nuclear matter. The success of these applications marks an important step toward applying RKDG methods to more realistic CCSN simulations with thornado in the future.

Nuclear astrophysics (1129)↗
DOE OSTI2022

Dry in-pile fracture test (DRIFT) for separate-effects validation of ceramic fuel fracture models

Fracture is an important component of nuclear fuel behavior, and significant efforts have been invested into developing fuel performance models that are capable of accurately representing fracture. Usable data on the process of fracture propagation in nuclear fuel under realistic service conditions are very limited. To address this need, a series of separate-effects experiments were developed and performed at Idaho National Laboratory's Transient Reactor Test (TREAT) facility. These experiments employ a heat sink to radially remove heat from the fuel in a manner that approximates the effect of coolant in an operating light-water reactor (LWR). The test holder for these experiments is known as the Dry In-pile Fracture Test (DRIFT). A series of experiments employing DRIFT and TREAT were performed to provide data on the extent and nature of fracture in fresh fuel at various points during a ramp to full power. Novel aspects of these experiments include the way they employ a heat sink to replicate steady-state LWR conditions, as well as the use of fiber optic sensors for in-reactor thermal instrumentation. Details on the development of this experiment, experimental conditions, and resulting data (including in situ thermal measurements and post-irradiation imaging of fracture) are provided in this work. LWR-equivalent powers ranging from 10 to 25 kW/m were tested using this apparatus. Cracking was visible at all power levels, with increasing cracking extent as the power level increased, although there was little difference in the cracking between the two highest-power tests, which had LWR-equivalent powers of 20 and 25 kW/m.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Corrosion-Resistant Coatings on Spent Nuclear Fuel Canisters to Mitigate and Repair Potential Stress Corrosion Cracking (FY23 Status)

This report summarizes the activities performed by Sandia National Laboratories in FY23 to identify and test coating materials for the prevention, mitigation, and/or repair of potential chloride-induced stress corrosion cracking in spent nuclear fuel dry storage canisters. This work continues efforts by Sandia National Laboratories that are summarized in previous reports from FY20 through FY22 on the same topic. In FY23, Sandia National Laboratories, in collaboration with five industry partners through a memorandum of understanding, evaluated the physical, mechanical, and corrosion-resistance properties of eight different coating systems. The evaluation included thermal and radiation environments relevant to various time periods of storage for spent nuclear fuel canisters. The coating systems include polymeric (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin, epoxy), organic/inorganic ceramic hybrids (silane-based polyurethane hybrid and a quasi-ceramic sol-gel polyurethane hybrid), and coatings utilizing a Zn-rich primer applied to stainless steel coupons. The results and implications of these tests are summarized in this report. These analyses will be used to identify the most effective coatings for potential use on spent nuclear fuel dry storage canisters and to identify specific needs for further optimization of coating technologies for application on spent nuclear fuel canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

NSRD-20, Functional Testing of Novel MTC HEPA Filtration Media

To improve the safety of Department of Energy (DOE) nuclear facilities, this project aimed to develop ceramic high efficiency particulate air (HEPA) filters that also create a low pressure-drop (dP) in use. The filters under consideration use mini-tubular ceramic (MTC) HEPA filtration media made with different architectures. Once successful, this project will reduce life-cycle costs, including safety basis, operational, and waste-disposal costs. The technology’s mini-tubular geometry is known to reduce dP when compared to flow through membranes of equivalent mass and surface area. Low dP facilitates retrofitting advanced filters into existing DOE facilities to make them safer while simultaneously reducing operational costs. Operations of ventilation systems are a key cost driver for DOE nuclear facilities; reducing dP can significantly reduce DOE’s operational costs. However, the filtration efficiency (FE) of this type of filtration media depends on the architecture of the filter media in an unknown way and requires testing. Functional testing of filtration media with different architectures will establish the performance of the filtration media and will enable development of guidance on design requirements to optimize filter performance.

36 MATERIALS SCIENCE↗
DOE OSTI2019

STEAM OXIDATION, BURST AND CRITICAL HEAT FLUX TESTING OF COMMERCIAL FECRAL CLADDING

A new accident tolerant Fe-12Cr-6Al-2Mo-0.2Si-0.03Y (C26M) fuel cladding was developed in response to the nuclear accident in Japan. Now that large quantities of commercial tubing are being fabricated, integral testing has expanded to include isothermal and non-isothermal steam oxidation testing, loss of coolant accident (LOCA) burst testing and critical heat flux (CHF) behavior. The oxidation behavior is being compared to other FeCrAl compositions and includes protective alumina scale formation during ramp testing at up to 11°C/min to 1475°C. Using standard LOCA conditions of 5°C/s heating in steam, burst testing was conducted with hoop stresses of 20-110 MPa and indicated higher burst temperatures for C26M compared to first generation FeCrAlY compositions without Mo and Si. Using the same LOCA equipment, a new experiment is being developed to simulate CHF conditions and the deformation behavior of C26M was compared to Zircaloy-4.

Pint, Bruce↗
DOE OSTI2022

Modeling Activities Related to Waste Form Degradation: Progress Report

This report represents the milestone deliverable M4SF-22SN010309092 “Modeling Activities Related to Waste Form Degradation: Progress Report” that describes the progress of R&D activities of ongoing modeling investigations specifically on nuclear waste glass degradation, Density Functional Theory (DFT) studies on clarkeite structure and stability, and electrochemical model development of spent nuclear fuel (SNF). These activities are part of the Waste Form Testing, Modeling, and Performance work package at Sandia National Laboratories (SNL). This work package is part of the “Inventory and Waste Form Characteristics and Performance” control account that includes various experimental and modeling activities on nuclear waste degradation conducted at Oak Ridge National Laboratory (ORNL), SNL, Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL).

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

Risk-informed Graded Approach for Reliability and Performance Assessment of Machine Learning and Artificial Intelligence for Advanced Condition Monitoring Techniques

With the shift away from time-based maintenance and toward condition-based maintenance, and to reduce overall maintenance costs, there has been an upsurge in the usage and development of advanced condition monitoring (ACM) techniques for real-time monitoring of nuclear power plant (NPP) components. ACM is particularly useful in the development of digital twins, which are designed to predict the failure or degradation of plant components. Successful implementation of ACM requires an assessment to inform the development of a risk-informed approach to evaluate the use of ACM to meet Nuclear Regulatory Committee (NRC) regulations for in-service testing (IST) programs. This includes the monitoring and diagnostics of reactor components and systems in current, new, and advanced reactors. A key component in ACM is the usage of machine learning (ML) and artificial intelligence (AI) algorithms that can employ real-time data from instrumentation and sensors to detect and predict reactor component degradations. Such predictive capabilities enable early detection of component degradation so as to help plant personnel plan and execute necessary maintenance. For successful implementation of ML/AI in ACM such that regulatory requirements are met, a risk-informed graded approach is needed to assess the reliability and performance of ML/AI for ACM. The American Society for Mechanical Engineers (ASME) developed their Operations and Maintenance (O&M) Code to provide guidance on safe, reliable O&M of NPPs. The IST section of the O&M Code specifically establishes requirements for IST and examination to gauge operational readiness of components in water-cooled NPPs. This paper presents a state-of-the-art review of how reliability and risk assessment can be integrated with ACM to assess component performance by non-nuclear industries. This is followed by different methodologies and approaches for conducting performance and reliability assessments so as to meet IST requirements for NPP components.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Wireless Instrumented RB Experiment Preliminary Design and Analysis

The ability to deploy new nuclear fuels for current or future reactor concepts requires carefully designed experiments to generate data to support fuel qualification. Ideally these experiments would include state of-the-art sensing to maximize the amount of in situ data that can be collected during operation. Furthermore, advanced reactor systems can take advantage of integrated in-core sensing technologies to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator’s understanding of limiting peaking factors. Before any novel sensing technologies can be readily adopted for nuclear applications, they must first demonstrate acceptable performance in test reactors. This report summarizes the preliminary design and analysis of the most highly instrumented irradiation experiment ever performed in the removable beryllium (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Wireless Instrumented RB Experiment 2021 (WIRE-21) will test a wide range of sensors including wireless sensors being developed by Westinghouse Electric Company (WEC) that could provide in situ measurements of peak fuel temperatures and fuel rod pressurization due to fission gas release. The ability to wirelessly transmit a signal through the fuel rod’s cladding is critical to improving fuel monitoring capabilities without requiring signal penetrations through the cladding pressure boundary, which would significantly impact fuel fabrication, handling, and operation. Other sensors that will be tested in WIRE-21 include an array of thermocouples, self-powered neutron detectors (SPNDs), and spatially distributed fiber-optic temperature sensors. More generally, WIRE-21 will establish a flexible irradiation vehicle design to allow accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. This report summarizes the mechanical design for WIRE-21, the experimental test matrix, initial neutronic and thermal design analyses, and the active monitoring and control system enhancements necessary to support testing of advanced sensor technologies. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) relevant to light water reactors (LWRs), but the flexible design of the experiment vehicle allows much higher operating temperatures (>1,100°C). Neutronic calculations determine the neutron flux conditions as well as the nuclear heating within the experiments. These results are used as inputs to detailed thermal finite element calculations, which are required to evaluate the complex, three-dimensional heat transfer that occurs within WEC’s wireless sensor enclosures. Initial results show that the temperatures of the sensors’ enclosures and the metal bellows can be operated near the temperature range of LWR coolants and cladding while simultaneously increasing the temperature of a surrogate fuel material to values in the range of 800–1200°C to simulate centerline fuel temperatures during LWR operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗