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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 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Application of Principal Component Analysis to Electrochemical Reprocessing PM and NMAC

In this report, data from an electrorefiner (ER) for nuclear fuel reprocessing is evaluated for process monitoring (PM) conclusions. This data comes from tests performed at the Idaho National Laboratory in 2022. Multivariate approaches utilizing methods of Principal Component Analysis (PCA) is applied. This is based off established work in process monitoring for fault detection in industrial facilities. This report will discuss the background, methods, and results of the application and some of the conclusions and applications that can be drawn from them. PCA is applied to two different electrorefiner (ER) operations that occurred at Idaho National Laboratory between August and October 2022. The first operation occurred with little incident while the second had several noted faults in the equipment in operational logs. The data from the first run was used to train the data for “normal” operations and applied to both sets of data to determine when operations were in an “off-normal” condition and identify where the fault occurs through PCA. PCA was able to identify off-normal events and identify the cause for off-normal operations. These identified off-normal events matched with the events and their causes in the operational logs. However, small amounts of variance in the data led to false detection of “off-normal” events. Thus, careful selection of training data and a-posteriori conclusions based off operator assessments will both be required for application of PCA to PM applications. This work demonstrated that multivariate approaches and latent variables are applicable to pyroprocessing PM applications and can be further expanded in future work as quality variables such as salt concentration from sensors and sampling become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Ceramic Seal.

Containment/Surveillance (C/S) measures are critical to any verification regime in order to maintain Continuity of Knowledge (CoK). The Ceramic Seal project is research into the next generation technologies to advance C/S, in particular improving security and efficiency. The Ceramic Seal is a small form factor loop seal with improved tamper-indication including a frangible seal body, tamper planes, external coatings, and electronic monitoring of the seal body integrity. It improves efficiency through a self-securing wire and in-situ verification with a handheld reader. Sandia National Laboratories (SNL) and Savannah River National Laboratory (SRNL), under sponsorship from the U.S. National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D), have previously designed and have now fabricated and tested Ceramic Seals. Tests have occurred at both SNL and SRNL, with different types of tests occurring at each facility. This interim report will describe the Ceramic Seal prototype, the design and development of a handheld standalone reader and an interface to a data acquisition system, fabrication of the seals, and results of initial testing.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2014

Application of Active Seismic and Electrical Methods to Detect and Characterize Subsurface Effects of an Underground Explosion

During the period of underground nuclear testing in the U.S., now ended for over 20 years, the focus of work was to conduct the test and analyze the data as quickly as possible; there was little concern about application of technologies that could be used to detect and characterize the location and depth of the explosion. Thus, there is little experience with applying geophysical methods in places where an underground nuclear explosion (UNE) has taken place. On the other hand, application of geophysical methods to the detection of shallow artifacts and water table depth is commonly applied in the environmental remediation and site characterization realm of commercial geophysical surveys. For nuclear nonproliferation treaty applications which may incorporate both shallow and deep geophysical probing methods (such as on-site inspection (OSI) under the Comprehensive Nuclear Test Ban Treaty [CTBT]), there thus remains a need to better understand detection capabilities of the deeper targets: anomalies related to an explosion cavity or rubble zone that may be located several hundred meters below the ground surface. The purpose of this Coordinated Inspection Technologies (CIT) project was to gain experience in applying active seismic methods and electrical methods at a site in Nevada where an underground nuclear test (UNE) had been conducted in the past to get a better understanding of the capabilities of these methods for detection and characterization of deep subsurface effects of the explosion.

58 GEOSCIENCES↗
DOE OSTI2019

Path to Subsurface Closure: When Real and Predicted Data Don't Match, Central Nevada Test Area (CNTA) - 19560

The Central Nevada Test Area, Nevada, was the site of an underground nuclear test in 1968, which resulted in residual contamination at and near the detonation depth of 975 meters (3200 feet). The environmental corrective action strategy for the site was conducted in accordance with the Federal Facility Agreement and Consent Order and all applicable Nevada Division of Environmental Protection (NDEP) policies and regulations. The corrective action strategy initiated in 1999 included developing a conceptual site model, preparing a numerical flow and transport model, calculating contaminant boundaries, negotiating compliance boundaries with NDEP, performing model validation, and monitoring groundwater. This strategy was executed through the Corrective Action Decision Document/Corrective Action Plan phase, when it was determined that the flow and transport model could not be validated because it failed to adequately predict groundwater elevations in the monitoring/validation (MV) wells (MV-1, MV-2, and MV-3) drilled subsequent to the modeling effort. This led to a change in the corrective action strategy that is now designed to validate the compliance boundary through monitoring and institutional controls, rather than relying predominantly on a numerical flow and transport model. This new approach included enhancements to the monitoring well network and additional monitoring to confirm that data were sufficient to proceed to closure. The corrective action activities were completed in 2015, and the Closure Report, which established the long-term postclosure monitoring program, was approved in 2016. The postclosure monitoring program is designed to (1) assess the effectiveness of the compliance boundary by monitoring for the radionuclides of interest and (2) evaluate the effectiveness of monitoring locations within the groundwater flow system by monitoring water elevations to ensure that monitoring wells are located along the potential migration pathways. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Performance of Waterlogged LWR fuel during RIA

Fuel pins with small defects could allow for the gradual ingress of water into the cladding during normal operation. If these 'waterlogged' fuel pins are subjected to a rapid power change, the internal pressurization could greatly exceed that of a normal fuel pin due to rapid conversion of the water to steam. This could induce either a premature or a more energetic failure during a design basis RIA than in non-defected fuel pins. Experimental studies on this type of fuel are very limited and do not appear to have been fully reported in the open literature. Results from a variety of tests conducted at the Transient Reactor Test Facility (TREAT), Special Power Excursion Reactor Test (SPERT) facility, Nuclear Safety Research Reactor (NSRR), and Power Burst Facility (PBF) are reviewed in this paper. These tests were intended to evaluate both the physical consequences of failure and the threshold and causes for waterlogged fuel failure during RIA. The available data suggests that although waterlogged fuel pins may burst at a lower energy deposition threshold, they do not pose a significant new hazard during design basis accidents. However, data is limited to tests conducted on fresh fuel samples only.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

The problem analysis of data processing performance in nuclear power plant computerized information and control system - 288

This paper is based on the data processing performance problems of operating command delay and system display not updating real-time for nuclear power plant computerized information and control system. The plant computerized information and control system is responsible for monitoring and controlling the plant operation, and it is a key part in the plant DCS system. Its data processing performance will affect the operation stability of nuclear power plant. The architecture, hardware configuration and data communication for plant computerized information and control system among the different levels in DCS are described. The main reasons affecting data processing performance in different signal channels are analyzed according to the plant load test results. The CPU load curves for real time server and central processing server are displayed, although the detailed cause for operating command delay and system display not updating realtime are not found, central processing server load is heavier than real time server which remind us to analyze the data processing of software resource scheduling deeply. So the resource distribution problem of software execution task is also analyzed according to CPU load test results. This paper provides the optimization analysis for data processing performance, and it is available and effective according to the latest test results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

MPACT Testing and Benchmarking Results

This document provides detailed reactivity and fission rate distribution results for MPACT for the Core Physics Benchmark Progression Problems 1-5, including comparisons to initial startup data from Watts Bar Nuclear Unit 1. Also included are computer requirements and runtimes for each type of problem and feedback for issues encountered during the testing. Testing for the benchmark problems that do not require thermal-hydraulic feedback has been very successful. Testing for the coupled problems (6 and 7) is still ongoing and will be documented in a future revision or separate document. A few issues have been identified needing more research/investigation before all of these problems have ideal solutions. These items are included in the Section 8 near the end of this document. MPACT has successfully solved all of the zero power core physics progression problems. The computer resources required and runtimes were all acceptable, though faster cases would be preferable considering there was no thermal-hydraulic feedback. MPACT converged very reliably for all 3D problems, expect in a few isolated studies where very thin axial planes were purposely created to attempt to quantify the errors in the control rod tip homogenization model. Despite the issues listed in Section 8, the accuracy of MPACT compared to the reference solution is considered to be very good.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

OSU Flow Test Fuel Plate Cladding -- Mechanical Properties Measurement. TEV-2105 R0

Tensile test specimens were made from shear drops from three such depleted-uranium fuel plates: CR6064; CS1027; and FP6686. Pieces for test-specimen fabrication originated from each end and along one side of all three plates. Specimen fabrication, testing, and analysis of resulting data are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Small-Scale Mechanical Testing of UO{sub 2}

The mechanical properties of UO{sub 2} fuel are of great importance to the nuclear engineering community for modeling of fuel performance during operation and accident scenarios. Data on the deformation of single crystals at room and elevated temperature can help these efforts by informing multiscale models. Small-scale testing techniques like micro-cantilever bending and nano-indentation have the ability to measure mechanical properties at the right length scale for this, while allowing the observation of the deformation of the material. In our work we perform micro-cantilever testing of UO{sub 2} at room temperature and nano-indentation testing up to 500 deg. C, as well as in-situ TEM micro-cantilever testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

ADVANCE System Requirements

In this document, we focus mainly on testing libraries written in the legacy ENDF format. As we move to new nuclear data format, we anticipate will need to revisit this document.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Subsurface Noble Gas Sampling Manual

The intent of this document is to provide information about best available approaches for performing subsurface soil gas sampling during an On Site Inspection or OSI. This information is based on field sampling experiments, computer simulations and data from the NA-22 Noble Gas Signature Experiment Test Bed at the Nevada Nuclear Security Site (NNSS). The approaches should optimize the gas concentration from the subsurface cavity or chimney regime while simultaneously minimizing the potential for atmospheric radioxenon and near-surface Argon-37 contamination. Where possible, we quantitatively assess differences in sampling practices for the same sets of environmental conditions. We recognize that all sampling scenarios cannot be addressed. However, if this document helps to inform the intuition of the reader about addressing the challenges resulting from the inevitable deviations from the scenario assumed here, it will have achieved its goal.

58 GEOSCIENCES↗
DOE OSTI2025

In-Cell Deployment and First Use of Digital Image Correlation for In-Situ Strain Analysis of Irradiated Nuclear Fuel Rods During LOCA Transient

Digital image correlation (DIC) is a noncontact, optical method increasingly used across industries and research environments for acquiring multidimensional strain data. At Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS), DIC has been extensively applied to study the thermomechanical response of nuclear fuel claddings, yielding fundamental insights into material behavior. However, these efforts have focused exclusively on unirradiated materials and relied on the SATS out-of-cell infrastructure. Efforts over the past two years have been made to extend these capabilities to ORNL’s Irradiated Fuels Examination Laboratory SATS system within the hot-cells to enable testing of irradiated fuel cladding materials. Integrating DIC into the hot-cell SATS infrastructure presents unique challenges, including enabling remote operation of optical equipment and adapting auxiliary hot-cell systems for DIC implementation. This report discusses design, stand-up testing, and application of DIC to an irradiated nuclear fuel cladding segment. A DIC testing rig was successfully built and validated out-of-cell through extensive surrogate tests and calibrations. The rig and specialized DIC furnace were installed in the hot-cell, and a DIC test was successfully conducted. Results from that test correspond to expectations for Zr-based alloys and showed similar uncertainties compared to out-of-cell tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2024

Innovative control mechanism for research and test reactors using mandrel-shaped control rods

Research and test reactors have historically played a pivotal role in supporting the initial development of nuclear reactors. They continue to provide essential data for enhancing fuel designs and material knowledge. However, with many such reactors aging and the growing demand for data to bolster advanced reactor development, it is more necessary to research potential design attributes of the next generation of research and test reactors. For test reactors dedicated to fuel and material testing, the design of control mechanisms significantly influences the stabilization of neutron flux levels in irradiation positions while sustaining criticality. This study presents an innovative control mechanism for potential research and test reactor designs. It employs small absorber rods that move in opposite axial directions to maintain axial symmetry of power and neutron flux during burnup cycles. These rods maximize reactivity worth while also offering flexibility to flatten the radial power distribution. An axial translation of the control mechanisms’ absorbers, as compared to the rotational movement of absorbers in control cylinders, also provides a benefit to available excess reactivity and cycle length. Additionally, this work utilizes a simplified core model of the Advanced Test Reactor to assess the performance of this control mechanism. Compared to the current control system based on rotating control cylinders, the new control mechanism has the potential to enhance, or at least maintain, neutronic performance parameters in this reactor design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Development of Mockups and Instrumentation for Spent Fuel Drying Tests

The purpose of this report is to provide updates on the experimental components, methodology, and instrumentation under development for use in advanced studies of realistic drying operations conducted on surrogate spent nuclear fuel. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. Water remaining in canisters upon completion of drying procedures can lead to cladding corrosion, embrittlement, and breaching, as well as fuel degradation. Additional information is needed on the drying process efficacy to help evaluate the potential impacts of water retention on extended longterm dry storage. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations. Smaller-scale tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report describes the implementation of moisture monitoring equipment on a pressurized, submersible system employing a single waterproof, electrically heated spent fuel rod simulator as a demonstration of analytical capabilities during a drying process. A mass spectrometer with specially designed inlets was used to monitor moisture and other gases at 150 kPa to 800 kPa for a test simulating a forced helium dehydration procedure and below 1 torr for tests mimicking a vacuum drying process. The dew point data from the mass spectrometer was found to be in good agreement with a solid-state moisture probe. A distinct advantage of the mass spectrometer system was the capability to directly sample from the hightemperature (>200 °C) head space expected in a prototypic scale experiment where a solid-state moisture probe would suffer considerable loss of accuracy or fail altogether. The operational and analytical experiences gained from this test series are poised to support an expansion to assembly-scale tests at prototypic length. These assemblies are designed to feature prototypic assembly hardware, advanced diagnostics for in situ internal rod pressure monitoring, and failed fuel rod simulators with engineered cladding defects to challenge the drying system with waterlogged fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets to the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗