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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 1,711 records · Page 95
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Theory and Implementation of Nuclear Safety System Codes - Part II: System Code Closure Relations, Validation, and Limitations

This is Part II of two articles describing the details of thermal-hydraulic sys- tem codes. In this second part of the article series, the system code closure relationships (used to model thermal and mechanical non-equilibrium and the coupling of the phases) for the governing equations are discussed and evaluated. These include several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models be- tween the fields. These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models. Code verification and validation, including Separate Effects Tests (SETs) and Integral effects tests (IETs) is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities, but some of the more advanced reactor designs will push the limits of validation for the codes. Lastly, the limitations of the codes are discussed by considering next generation power plants, such as Small Modular Reactors (SMRs), analyz- ing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, in- novative fuel types, and many other design features that may not be fully analyzed by current system codes. This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations, val- idation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

An analysis method for experiment power in the ATR

The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) is used for testing and qualification of nuclear fuels and materials. The ATR has been irradiating the lead test assembly (LTA) for the planned Ki-Jang Research Reactor (KJRR) in Korea. Irradiation and subsequent Post-Irradiation Examination (PIE) of the LTA will lead to qualification of the fuel to support licensing activities. Qualification of the LTA requires the experiment to meet specific thermal and burn-up conditions. In particular, plate 20 of the experiment must obtain a heat flux between 137 and 200 W/cm{sup 2} with a target U-235 depletion of 79-84%. Due to the limited instrumentation and burn-up monitoring capabilities, assurance of meeting experiment goals is verified through modelling. In order to validate the experiment model using ATR operating data, a methodology was developed to compare the calculated experiment power with ATR power monitoring instrumentation. This paper discusses the methodology used to quantify the experiment power and uncertainty in the model based on ATR operating conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

MEXIICO: a new equipment to study the pressure impact on the irradiated fuel behavior

The new MEXIICO experimental loop has recently been started in the LECA-STAR hot-cell facility of the CEA Cadarache. It is dedicated to the understanding of the basic mechanisms occurring in nuclear fuels during nuclear reactor power transients and accidental situations. Under such conditions the nuclear fuels are submitted to thermo-mechanical stresses. These stresses play a major role on fission product release and fuel fragmentation. Evaluating such fuel behavior represents a key issue for reactor safety analyses. In the first part of the paper, we describe the complete MEXIICO device, which main component is a hot isostatic press allowing to perform analytical experiments in the [20 -1600 C. degrees / 10 MPa - 160 MPa] region. As shown by simulations (ALCYONE fuel performance code), the upper value of the pressure (160 MPa) is consistent with the one experienced by fuels under nuclear power transients. We also present the experimental loop starting from the argon pressurized feed-line, the furnace and its associated instrumentation and the exhaust line equipped with specific valves and flowmeters controlling the pressure and flowrate. We then focus on the gamma spectrometry measurement which allows recording the fission gas release kinetics. In the second part of the paper, we detail the capability of the MEXIICO device to provide new experimental data regarding the influence of pressure on the fission gas release and fuel fragmentation, and to determine a pressure threshold above which fission gas release and fuel fragmentation may be inhibited. This device also allows investigating the impact of pressure on the microstructure evolution, coupling it with post-test microanalysis examinations: this procedure will allow characterizing the effect of temperature and pressure on the bubble evolution and interconnection. Evolutions of the fuel modeling, needed to simulate correctly these new experimental data, will be also presented. In the last section of the paper, we present the commissioning tests of the MEXIICO device showing the good reproducibility of the tests performed first in conventional laboratory and then in hot-cell laboratory. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Thermal Hydraulics Validation Activities in the ART-GCR Campaign

The Advanced Reactor Technologies - Gas Cooled Reactor Campaign (ART-GCR) is part of the United States Department of Energy - Office of Nuclear Energy's Advanced Reactor Technologies program aimed at developing a High Temperature Gas-cooled Reactor (HTGR) which will offer enhancements in safety and efficiency. The Design, Methods, and Validation arm of the ART-GCR program provides experience and advanced tools for HTGR design and analysis. The Methods and Validation arm is developing data to validate predictions of decay heat removal through an extensive experimental campaign the Natural Convection Shutdown Heat Removal Test Facility at Argonne National Laboratory. In-core thermal hydraulics validation is being facilitated through the OECD/NEA's Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data. That benchmark contains code-to-code and code-to-data comparisons for 3 sets of HTGR thermal hydraulics phenomena. This ongoing benchmark activity has 16 participants from 9 countries developing leading insights into HTGR thermal hydraulics code validation. The ART-GCR campaign is also working on the development of state-of-the-art MOOSE-based models of the High Temperature Engineering Test Reactor (HTTR). Through the OECD/NEA's HTTR Loss of Forced Cooling (LOFC) program, the ART-GCR campaign is developing a multiphysics transient code validation benchmark based on 3 HTTR experiments that will provide an opportunity for international collaboration on multiphysics code validation.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

MRI of testicular malignancies

Although testicular carcinoma represents approximately only 1% of solid neoplasms in men, it is the most common malignancy between young men. The two main histologic categories are testicular germ cell tumors (TGCTs), including seminomas and nonseminomas, accounting for 90–95% of testicular neoplasms and sex cord-stromal tumors. Scrotal MRI, including a multiparametric protocol, has been proposed as a valuable supplemental imaging technique in the investigation of testicular pathology. Recently, the Scrotal and Penile Imaging Working Group appointed by the board of the European Society of Urogenital Radiology has produced recommendations on when to perform scrotal MRI. Regarding intratesticular masses, MRI of the scrotum may be used for their characterization, when US findings are indeterminate and for local staging of TGCTs, when organ-sparing surgery is planned. Differentiation between seminomas and nonseminomas is possible based on MRI features, when clinically needed. Scrotal MRI may also help in differentiating between TGCTs and nongerm cell tumors. Functional information based on diffusion-weighted imaging and dynamic contrast-enhanced MRI data improve testicular mass lesion characterization. Preliminary observations on diffusion tensor imaging, magnetization transfer imaging, and proton MR spectroscopy bring about new data in the understanding of testicular microstructure and pathophysiology.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Force Reconstruction Using the Inverse of the Mode-Shape Matrix [Book Chapter]

A new technique for force reconstruction is developed. To estimate the externally applied force, this technique sums the weight-scaled acceleration signals, and is referred to as the Sum of Weighted Accelerations Technique (SWAT). To obtain the scalar weights the inverse of the mode shape matrix is used. Application of this technique is illustrated with both numerical calculations using a mass-spring model and experimental data from a structure impacting a rigid barrier.

055001 -- Nuclear Fuels-- Safeguards, Inspection, ↗
DOE OSTI2017

Identification and repair of intermittent cable faults in nuclear power plants - 131

The nuclear industry uses cable circuits as an integral part of a plant's reactor protection system (RPS), instrumentation and control (I and C) and power distribution systems. The reliability of these systems is critical for continued operation of the world's aging nuclear fleet, as well as for safe shutdown of the reactor. When intermittent faults occur in cable circuits, the symptom of the problem often appears and disappears in milliseconds while the aftereffects can result in spurious alarms and automatic actuation of safety systems. Problems are compounded when component degradation causes a cable circuit to transmit inaccurate information that does not represent actual plant operating conditions. The root cause of intermittent cable circuit operation can include degradation of cable and connectors from heat, humidity, radiation, and vibration. Unlike hard faults (which are constant open or short circuits), intermittent faults are defined as a temporary change in impedance that may only last for a few milliseconds. While less disruptive than hard faults, signal spiking from intermittent faults can often lead to alarm conditions and reactor trips. A soft fault is defined as a subtle variation the electrical characteristics of the circuit that still permit the continued operation of the circuit. A soft fault may be a precursor to a hard fault due to continued degradation of the fault over time. The focus of this paper is the description of a cable testing system that can be connected to an active, low frequency circuit and continuously monitor electrical characteristics using the time domain reflectometry (TDR) technique until an intermittent impedance change occurs. This technique uses the difference in frequency response of the circuit and the TDR to isolate the TDR signature from the active circuit data . When a momentary impedance change occurs that exceeds a preconfigured limit threshold, the distance to fault data is automatically overlaid with the healthy baseline reference and stored in a database for diagnosis and repair. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Real-Time Characterization of Salt Aerosols Generated from Static and Sparged Molten Salt

The formation of radionuclide-bearing aerosols in the respirable size range has the potential to significantly influence offsite dose consequences and is, therefore, an important consideration in nuclear facility safety assessments. Molten salt reactor (MSR) developers will likely need to demonstrate an understanding of the conditions under which radionuclide-bearing aerosols may be generated from their reactor under normal operating and accident conditions, as well as the characteristics and transport behavior of these aerosols, to demonstrate to the U.S. Nuclear Regulatory Commission (NRC) that the facility can be operated safely. Recent reviews of the literature identified a lack of experimental data describing the mechanisms of formation and properties (size, concentration, and composition) of salt aerosol particles that are produced from molten salts. Experiments that identify the conditions that lead to radionuclide-bearing salt aerosol releases and quantify the characteristics of salt aerosols formed by different mechanisms are high-priority needs to support MSR licensing. This report describes tests that were conducted within the Argonne Salt Aerosol Test Stand (a sealed vessel and measurement system) to generate salt aerosols from static and sparged molten salts and measure their size and concentration in real-time. The results provide insight into salt aerosol formation by the vapor condensation and bubble bursting mechanisms and inform the potential radiological consequences of aerosol formation from molten fuel salt. Videos of the salt surface were taken during salt sparge tests to observe surface bubble behavior. The data in this report can be used to develop mechanistic source term and accident progression models for MSRs. The real-time salt aerosol characterization technique used in this study will be employed in future integral effects tests that are conducted at an engineering scale to simulate realistic MSR accidents and in future separate effects tests to address additional variables that may impact salt aerosol characteristics (e.g., presence of fission products in salt and humidity in atmosphere).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Leveraging Machine Learning Capabilities for the Characterization of Irradiated Uranium: A Case Study of Analysis Methods for Nuclear Safeguards and Nuclear Forensics

Nondestructively determining the initial enrichment of irradiated uranium is a complex and laborious multivariable problem due to the presence of fission products. This work demonstrates the capabilities of machine learning to analyze gamma-ray spectral data to determine initial enrichment without knowledge of the decay time of the sample. The approach developed is agnostic to the particular scenario and is applicable to a wide variety of applications in nuclear forensics and nuclear safeguards. We irradiated 5 mg uranium standard reference materials at discrete enrichment values ranging from 0.02% to 97% 235 U (weight percent) in UT Austin’s Nuclear Engineering Teaching Laboratory TRIGA Mark II 1.1 MW research reactor, allowed each to decay for 8 hours, and then measured each sample via gamma-ray spectrometry for 50 hours post-irradiation yielding 1,400 individual gamma-ray spectra discretized into 8,192 energy bins. We then trained decision trees models to analyze individual gamma-ray spectra and estimate the associated initial enrichment without knowledge of the time since end of irradiation. We evaluated the performance of the models with a reserved test set not used for training or calibrating the model. A decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 2.3% (weight percent 235 U). Next, we implemented a principal component analysis pre-processing routine of the gamma-ray spectrometry data to reduce the dimensionality of the dataset from 8,192 channels in the spectrum to 10 principal components while retaining over 99% of the inherent variance in the data. Decision tree models constructed with these data demonstrated decreased mean absolute error in enrichment determination, reduced computational time, and decreased complexity. A single decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 0.05% (weight percent 235 U). Furthermore, we analyzed these models with learning curves to ensure that overfitting did not occur. The capabilities provided by these models can be naturally extended to other application-focused measurements in the fields of nuclear safeguards, nuclear forensics, and nuclear non-proliferation.

Drescher, Adam↗
DOE OSTI2020

Neural Learning Based Blind Source Separation for Detection of Material Defects in Pulsed Thermography Images

In this paper, we introduce a neural learning-based approach to blind source separation for detection of material flaws in pulsed thermography (PT) images. This approach can be used to detect internal defects (pores) in metallic Additively Manufactured (AM) materials. Such defects occur in high-strength alloys produced with direct laser sintering AM method for nuclear energy applications. Pulsed thermal imaging system utilizes a high intensity flash lamp to rapidly heat surface of sample, a high sensitivity infrared camera to capture data of surface temperature variations. The data cube obtained with PT (stack of surface temperature images at different times) can be analyzed with image processing algorithms to detect material defects. Compared with conventional nondestructive evaluation (NDE) methods, such as digital radiography and ultrasonic testing, the PT pulsed infrared thermal detection technique has advantages of one-sided non-contact measurements, fast processing of large sample areas captured in one image. Detection of small material defects requires finding features in thermal images, 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 the photothermic, the NDE system detects internal calibrated defects of various sizes and depths in AM nuclear-grade metallic alloys.

Zhang, Xin↗
DOE OSTI2019

Advanced Retrieval Demonstrations of the Hanford Waste End Effector - 19383

This paper describes advanced retrieval demonstrations of the Hanford Waste End Effector (HWEE) at the Washington River Protection Solutions (WRPS) Cold Test Facility conducted by Pacific Northwest National Laboratory (PNNL) in co-operation with the U.S. Department of Energy - Office of River Protection and WRPS. The HWEE is a water-based scarifier originally developed by PNNL for the U.S. Department of Energy tank retrieval systems in the 1990's as the confined sluicing end effector and later used with additional development in the waste retrieval of tanks at Oak Ridge National Laboratory. The HWEE is being assessed for potential use in Hanford tank retrievals because it is expected to use less water during waste retrieval than other water-based methods (e.g., sluicing) previously used in Hanford tanks. This is primarily because the HWEE uses three rotating high-pressure jets for dislodging the waste and a high-velocity conveyance system to retrieve the water and the waste locally at the point of waste dislodging. The retrieval demonstrations build on previous development testing (Phase I) completed in 2017 by addressing key tank waste challenges, including retrieval of a mixed layered simulant and retrieval of uneven simulant profiles like that found near tank walls. The tests involved retrieval of water and kaolin (sludge) simulant to determine pumping rate, dilution factors, and a demonstration of vertical pumping capability. Upgrades to the HWEE for Phase II testing included a more robust inlet screen, internal flush jet modification for lower water usage, and removal of the external flush mole that was not used during Phase I retrieval testing. In addition, a rotational positioning capability was added to the three-dimensional gantry to gain additional movement capability and flexibility needed to implement more complex retrieval mining techniques. Two simulant configuration options using existing simulant materials were defined to represent key Hanford tank waste properties: 1) layered waste (e.g., strong crust over weaker sediment) and 2) topography that is representative of tank waste (uneven surface). Historical information from Hanford waste, including tank history, sampling reports, and images, was investigated to define the simulant characteristics and physical configuration. Using the layered simulant configuration, retrieval tests focused on demonstrating the same operational parameters irrespective of simulant strength. The focus of the topographical retrieval test was to demonstrate the ability to clean next to the tank wall using a tilted HWEE and to demonstrate water confinement using a specific mining strategy. This paper will describe the Phase II test equipment, instrumentation, simulant preparation procedures, and testing methodology. Testing results that will be presented include HWEE water usage versus operating pressure, simulant characterization data, dilution ratios, and retrieval rates. Reaction forces and other factors that would affect the remote operation of the HWEE will also be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Multiple Flow Loop SCADA System Implemented on the Production Prototype Loop

The following report covers FY 15 activities to develop supervisory control and data acquisition (SCADA) system for the Northstar Moly99 production prototype gas flow loop. The goal of this effort is to expand the existing system to include a second flow loop with a larger production-sized blower. Besides testing the larger blower, this system will demonstrate the scalability of our solution to multiple flow loops.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

OSIRIS-REx, bringing back a bit of an asteroid

The OSIRIS-REx mission will provide invaluable ground-truth validation for the computer models in a case study that set out to determine the timeframe within which we must launch either a kinetic impactor or a nuclear explosive carrier and develop a conceptual design for that spacecraft. From the on-asteroid data gathered when the spacecraft touches down and, eventually, from hands-on research here on Earth, we will get a reality check for what we think we know about Bennu’s material composition (boulders, chunks, and dust), shape, mass, density, and so on. Gravimetric data could reveal whether the asteroid is hollow. Back on Earth, researchers will run various tests on the material. We may have enough information to develop a theoretical equation of state. We could then compare it to its closest match in SESAME, the Los Alamos database of equations of state. Armed with that data, we will rerun the models to see how much each scenario changes. Comparing data from the touchdown and sample return to the earlier, remotely observed data from Bennu will help us calibrate remote observations of other objects in the future. Then we can more precisely model the effects of our deflection and disruption strategies and refine the spacecraft design.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Improving to the neutron fluence rate monitor measurement system at the Advanced Test Reactor

The existing fluence monitor wire scanning system at the Advanced Test Reactor (ATR) was designed and installed for use in the Engineering Test Reactor (ETR) when it began operation in 1958. The wire scanner was operated in ETR for over 20 years until ATR began operation, when it was moved to the ATR west canal area in 1971 and subsequently moved to the west canal in 2006 where it presently resides. With a continued service life of 65 years the system is well beyond the typical design life of 20 years for these types of systems. The need to update the data acquisition and control system was identified, and the benefits of replacing the existing sodium iodide (NaI) detector with an electronically cooled high-purity germanium (HPGe) detector are discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. In this work, Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Search for a Light Sterile Neutrino at Daya Bay

In this study, a search for light sterile neutrino mixing was performed with the first 217 days of data from the Daya Bay Reactor Antineutrino Experiment. The experiment’s unique configuration of multiple baselines from six 2.9 GW th nuclear reactors to six antineutrino detectors deployed in two near (effective baselines 512 m and 561 m) and one far (1579 m) underground experimental halls makes it possible to test for oscillations to a fourth (sterile) neutrino in the 10 - 3 eV 2 < | Δ m 41 2 | < 0.3 eV 2 range. The relative spectral distortion due to the disappearance of electron antineutrinos was found to be consistent with that of the three-flavor oscillation model. The derived limits on sin 2 2 θ 14 cover the 10 - 3 eV 2 ≲ | Δ m 41 2 | ≲ 0.1 eV 2 region, which was largely unexplored.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
NASA NTRS1981

Evaluation of HCMM satellite data for estuarine tidal circulation patterns and thermal inertia soil moisture measurements

Digital thermal maps of the Cooper River (SC) and the Potomac River estuaries were prepared from heat capacity mapping radiometer (HCMR) tapes. Tidal phases were correctly interpreted and verified. Synoptic surface circulation patterns were charted by location thermal fronts and water mass boundaries within the estuaries. Thermal anomalies were detected adjacent of a conventional power plant on the Potomac. Under optimum conditions, estuaries as small as the Cooper River can be monitored for generalized thermal/tidal circulation patterns by the HCMM-type IR sensors. The HCMM thermal inertia approach to estimating soil moisture at the Luverne (MN) test site was found to be unsatisfactory as a NESS operational satellite technique because of cloud cover interference. Thermal-IR data show similar structure of the Baltimore and Washington heat islands when compared to NOAA AVHRR thermal-IR data. Thermal anomalies from the warm water discharge water of a nuclear power plant were mapped in Lake Anna, Virginia.

Wiesnet, D. R.↗