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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Constraining the Charge-Sign and Rigidity-Dependence of Solar Modulation

Our ability to identify the sources of cosmic rays and understand how these particles propagate through the interstellar medium is hindered by the combined effects of the solar wind and its embedded magnetic field, collectively known as solar modulation. In this paper, we build upon our previous work to model and constrain the effects of solar modulation on the cosmic-ray spectrum, using data from AMS-02 and BESS Polar II collected between 2007 and 2012, during which the heliospheric magnetic field was in a state of negative polarity. Our model uses measurements of the heliospheric magnetic field and the tilt angle of the heliospheric current sheet to accurately predict the effects of solar modulation as a function of time, charge, and rigidity. By incorporating data from a period of negative polarity, we have been able to robustly observe and constrain the charge-dependent effects of solar modulation.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2020

Micro-imaging with an atomic magnetometer and flux guide

Micro-imaging may be performed with an ultra-sensitive atomic magnetometer (AM) and an array of flux guides (FGs). The array of FGs may be configured to act as a magnetic lens that expands microscopic magnetic distribution to match dimensions of the AM. A plurality of single channel AMs may be combined into an array, or the AM may include an array of photodetectors, to realize multi-channel operation.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

Aerial Measuring System - Analysis of the Releasable Data Set

This is a collection of spectral data files obtained from AMS flights over areas in and around Las Vegas and the Nevada National Security Site. These are being made available for anyone to analyze and compare results with those in the accompanying report. The data files are attached to the PDF as Excel CSV format files.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Snowmass2021 Cosmic Frontier White Paper: Puzzling Excesses in Dark Matter Searches and How to Resolve Them

Intriguing signals with excesses over expected backgrounds have been observed in many astrophysical and terrestrial settings, which could potentially have a dark matter origin. Astrophysical excesses include the Galactic Center GeV gamma-ray excess detected by the Fermi Gamma-Ray Space Telescope, the AMS antiproton and positron excesses, and the 511 and 3.5 keV X-ray lines. Direct detection excesses include the DAMA/LIBRA annual modulation signal, the XENON1T excess, and low-threshold excesses in solid state detectors. We discuss avenues to resolve these excesses, with actions the field can take over the next several years.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2013

A Review of Tungsten Heavy Alloy Utilization in Isotope Transport Containers - 13380

A common requirement for radioisotope transport containers is that they provide both durable and efficient shielding of penetrating gamma radiation. This is the case for transport of both spent nuclear fuel as well as intentionally created radioisotopes for medical or other uses. Tungsten heavy alloy (WHA) provides a unique engineering property set for such shielding - easily surpassing more commonly used lead alloys in both strength and attenuation. This family of alloys contains typically 90-98 wt.% W in combination with transition metals such as Ni and Fe. WHA is manufactured in near net shape blanks by liquid phase sintering of compacted powder shapes to full metallurgical density parts. This powder metallurgy approach is described in its ability to provide excellent material utilization and affords efficient manufacturing of various shapes required for gamma shields or collimators. WHAs offer very high density (approaching 19 g/cc) in combination with relatively high thermal conductivity, low thermal expansion, ambient corrosion resistance, and can be provided with mechanical properties comparable to many medium carbon steels. As such, they can be machined to complex, damage resistant geometries using common metal cutting tools and methods. WHA additionally provides a lower toxicity alternative to Pb- or U-based gamma shielding. Given the specialty nature of WHA, specific metallurgical characteristics are reviewed to assist shielding designers who may otherwise encounter difficulties locating important alloy selection and fabrication details. Contained within this materials and applications overview are guidelines for WHA component design, alloy selection, and practical machining, finishing, and assembly considerations. The microstructure of WHA is that of a metal matrix composite. This factor has specific implications in the design of components for stress service as well as their protection in the presence of electrolytes. WHA is also discussed in the broader context of materials compatibility, as it is rarely used in isolated monolithic shapes. Alloy selection for new applications is often made primarily on the basis of density. An alternative strategy to this selection approach is presented which proposes that mechanical requirements for a given shielding use be the primary selection criterion over density. Standard commercial grades of WHA for radiation shielding are defined by specifications such as AMS-T-21014 and ASTM B777. These specifications define 4 density classes of as-sintered WHA. Compositional options, as well as post-sinter processing of WHA, are discussed for shielding components that must exhibit higher levels of ductility or very low magnetic permeability. In addition to the mechanical advantages over Pb-based shielding, the higher linear attenuation of energetic photons for various grades of WHA (as calculated by the NIST XCOM routine) are presented for selected photon energies of interest to illustrate the shield volume reduction generally possible through the use of tungsten-based shielding. While W provides inefficient attenuation of neutrons in a mixed radiation environment, its secondary role in shielding gamma radiation produced as a result of neutron capture is also described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Impact of lump characteristics on self-shielding corrections of active neutron interrogation on fissile material in waste - 15405

Non-destructive active neutron assay techniques such as active neutron coincidence counting, differential die away analysis or the californium shuffler method rely on the interrogation of fissile nuclear materials by an external neutron source. Whether it is a spontaneous fission or (α,n) neutron source, like {sup 252}Cf or AmLi; or a fusion neutron generator, all of these techniques rely on induced secondary fission events to provide prompt or delayed neutron signals that serve as the basis for the fissile material quantification. The accuracy of these fissile mass measurements is therefore dependent on the neutron interrogation efficiencies determined for calibration reference standards and whether the measured items are comparable to the calibration condition. In such favorable conditions, it is possible to establish a reliable empirical non-linear relationship between apparent (uncorrected) and real or true fissile mass. However, in waste measurements, this is unlikely to be the case; thus a calibration approach that takes account of the neutron self-shielding effect is required. By incorporating the Self-Shielding Factor (SSF) of the calibration reference standards, a linear mass calibration factor [s{sup -1}. {sup 239}Pu{sub eq} g{sup -1}] can be found; this calibration is accurate for waste in which the fissile material is 'infinitely dilute' i.e. the waste exhibits no self-shielding; this is assumed to be representative of dispersed material. The self-shielding correction and its systematic uncertainty are then left to be treated as part of the Total Measurement Uncertainty (TMU) evaluation associated with the assay result. The derivation of the TMU contribution associated with the self-shielding effect in nuclear waste is a complex problem for which a reliable evaluation has to be performed. This is usually based on assuming 'worst case scenarios' for the distribution of material that can arise in a particular waste stream. Such an approach is conservative, but often very penalizing, especially if one considers larger lumps (those containing grams of fissile material); for such lumps, the self-shielding correction becomes very significant, potentially altering the mass of {sup 239}Pu{sub eq} by over an order of magnitude. A worst case 'nuclear safety value' may not be the most appropriate choice for building a consignment best estimate or repository sentencing value for an item. In the present work the authors describe a numerical approach to study the variability of neutron self-shielding factors associated with distributions of a wide range of lump characteristics in terms of size, shape, chemical composition, density and isotopic composition. The described novel calculation tool is capable of accurately predicting the real impact of random distribution of lumps in a waste material, and allowed to show that given reasonable, defensible, assumptions on the composition of the lumps in waste stream, the systematic error component of the DDA TMU can be significantly reduced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Open-air Demolition: Validation of EPA Approved Alternative Method for Emission Estimation - 19397

Emissions from open-air demolition of US Department of Energy (DOE) facilities are regulated by the US Environmental Protection Agency (EPA) under 40 CFR 61 subpart H. This regulation requires that facilities estimate emissions using an EPA approved method from activities such as open-air demolition prior to commencing the activity. Should those emissions exceed certain thresholds, the facility may be required to obtain an Approval to Construct or Modify the facility and/or installed monitoring methods meeting methods acceptable to EPA. There are no EPA approved methods that are specific to open-air demolition methods. For this project the DOE and CH2M Hill BWXT elected to prepare an alternative method specific to open-air demolition. EPA conditionally approved the method but required validation of the method during the demolition of the Vitrification Facility at the West Valley Demonstration Project (WVDP) before use for open-air demolition at other WVDP structures. The validation method included detailed source characterization, close-in ambient samplers, and a comparison of measured results to predicted emissions using AERMOD (American Meteorological Society (AMS) and U.S. Environmental Protection Agency (EPA) Regulatory Model). The methods evaluated include hydraulic hammers, mechanical shears, load out of waste and handling of the debris piles. The validation method involved the collection and analysis of ambient air samples from air samplers located at strategic locations within the projected plume path of the demolition activities. The analytical results from the air samplers were then compared to ambient air concentrations predicted at the sampler locations with the EPA model AERMOD. The input to the AERMOD software is a source term developed weekly using the alternative method and material at risk disturbed during each week. To account for the building wake effects of the surrounding structures, virtual point sources were developed in the AERMOD model and placed at the locations of the demolition, debris piles, debris handling and load out areas. It was found that the factors used for hydraulic hammering were extremely conservative and that emissions from the debris pile were not only the primary emission source, but that the moisture content of the debris pile and the wind speed were critical factors that must be considered in future open-air demolition projects. Based on the results of the validation process, adjustments were made to the alternative method. The final validated alternative method provides an approach for estimating airborne emissions from open-air demolition methods and techniques. The method will be used in lieu of 40 CFR 61 Appendix D to estimate emissions from future open-air demolition projects at the WVDP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Verification of Special Nuclear Material (SNM) using Active Well Coincidence Counter (AWCC) and a High Purity Germanium (HPGe) Detector

SNM verification can be accomplished via two methods i.e. destructive analysis (DA) or nondestructive analysis (NDA). The DA modes are much more accurate compared to NDA modes; nevertheless, the sample's physical integrity is destroyed in the process. Techniques such as ICP-MS, ID-ICP-MS, TI-MS are few examples of DA, where a bulk material is ground and mixed with a solution (some form of acid) to acquire a homogenous sample. In contrast, NDA modes utilize signals emitted by the sample such as gamma, neutrons, light, or heat to analyze samples under investigation. HPGe detectors are some of the most frequently used gamma detectors to assay SNM. However, in a high activity scenario it provides a large uncertainty and sometimes becomes unfeasible to measure low energy gamma rays. Meanwhile, neutron detectors are not affected by the gamma background, therefore they can be used in addition to the gamma detectors to verify total fissile mass ({sup 235}U + {sup 239}Pu) of SNM. The AWCC is one of the thermal neutron NDA systems developed at the Los Alamos National Laboratory (LANL). It consists of 42 {sup 3}He detectors to measure gross neutrons (singles), coincidence neutrons (doubles), higher order multiplicity counting (triples, quadruples, etc.). It can effectively operate in both active and passive neutron interrogation modes. It can utilize AmLi, AmBe, or {sup 252}Cf active interrogation neutron sources. It is used to verify declared fissile mass ({sup 235}U or {sup 239}Pu) present in SNM. Objectives: Calibrate HPGe and AWCC for lightly irradiated (with enough {sup 137}Cs gamma background) ∼93% enriched HEU fuels; Verify fissile mass ({sup 235}U +{sup 239}Pu) content using AWCC and HPGe; Model and benchmark AWCC in MCNP6. Neutron coincidence and multiplicity counting: Two or more neutrons are coincident if they are detected by the system within the specified gate window; In AWCC: Gate window is 128 μsec; Induced fission in fissile material such as {sup 235}U and {sup 239}Pu - emission of time correlated neutrons - Singles, Doubles, Triples, or Quadruples; No induced fission in non-fissile materials- no correlated neutrons - only singles. SNM can be verified with high accuracy and precision using gamma and neutron NDA instruments such as HPGe and AWCC. Interaction with thermal neutrons. Non-fissile atoms - do not emit correlated neutrons. Fissile atoms - emit correlated neutrons during induced fission events. The time correlated neutrons can be detected by using coincidence or multiplicity counters such as AWCC. Likewise, {sup 235}U produces 186 keV gamma peak that can be measured by a HPGe gamma detector. Counts under 186 keV gamma peak provide information about {sup 235}U content. Combining AWCC and HPGe results can provide better understanding of special nuclear material under investigation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Emerging and existing sensing technologies for small modular reactors and advanced reactors - 191

The development of small modular reactors (SMRs) and Generation-IV advanced reactor designs represents a promising future for nuclear energy. These next-generation reactors can support new applications for nuclear energy like hydrogen production, industrial heat generation, and water desalination in addition to electricity generation. Furthermore, SMRs and advanced reactors enable load-following operations, offer improved passive safety features, support lengthened refueling cycles, and provide spent fuel management capabilities. To date, there are no SMRs operating or under construction in the United States, but commercial operation of the first integral Pressurized Water Reactor (iPWR) SMR is expected by the mid-2020's with Generation-IV advanced reactors to follow a decade or two later. To facilitate the development and deployment of SMRs and advanced reactors, key instrumentation and control (I and C) system challenges must be addressed and existing nuclear grade I and C sensors will need to be adapted or new sensors developed. For example, traditional measurement techniques and maintenance strategies used in the current generation of light water reactors (LWRs) will be challenged by the inherent design characteristics of the iPWR. For Generation-IV reactors, conventional I and C sensors will not withstand prolonged exposure to the harsh environmental conditions of advanced reactors (e.g. elevated temperatures, corrosive coolants, increased radiation). This paper discusses the challenges associated with making primary system process measurements in next-generation reactors and includes a summary of work performed to evaluate existing and emerging sensor technologies for these systems. As importantly, the paper presents the initial test results of a prototype temperature sensor developed by Analysis and Measurement Services Corporation (AMS) for use in SMRs and advanced reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

On-line monitoring of I and C transmitters and sensors for calibration verification and response time testing was successfully implemented at ATR - 326

The Advanced Test Reactor (ATR) has always had a comprehensive procedure to verify the performance of its critical transmitters and sensors, including RTDs, and pressure, level, and flow transmitters. These transmitters and sensors have been periodically tested for response time and calibration verification to ensure accuracy. With implementation of on-line monitoring techniques at ATR, the calibration verification and response time testing of these transmitters and sensors are verified remotely, automatically, hands off, include more portions of the system, and can be performed at almost any time during process operations. The work was done under a DOE funded SBIR project carried out by AMS. As a result, ATR is now able to save the manpower that has been spent over the years on manual calibration verification and response time testing of its temperature and pressure sensors and refocus those resources towards more equipment reliability needs. More importantly, implementation of OLM will help enhance the overall availability, safety, and efficiency. Together with equipment reliability programs of ATR, the integration of OLM will also help with I and C aging management goals of the Department of Energy and long-time operation of ATR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Application of quantitative methods for reliability testing of a nuclear power plant digital rod position indication diagnostic system - 391

Obsolescence, aging, reliability, and performance issues are driving nuclear facilities to replace conventional analog Instrumentation and Control (I and C) systems with digital technologies. In addition, the designs of the next generation of nuclear reactors, including Small Modular Reactors (SMRs), will incorporate digital I and C for most, if not all, of their safety and non-safety related functions. As part of an on-going research and development (R and D) effort under the auspices of the U.S. Department of Energy (DOE), Analysis and Measurement Services Corporation (AMS) is developing a platform to provide automated testing of digital I and C systems and create a standard method of evaluation for reliability assessments. This hands-on R and D effort has produced a Software Reliability Tester (SRT), which is a set of software and hardware tools designed to automate the testing of digital I and C systems to measure and quantify how well the systems perform under normal operating conditions and in the presence of faults. Furthermore, the SRT provides the foundation for a practical tool to automate verification and validation (V and V) activities and reduce the amount of testing time of digital I and C systems. When combined with its capabilities of integrating both reliability and fault tolerance quantification, the SRT can be used to ensure that digital I and C implementations are both safe and cost-effective for the nuclear industry. This paper describes the application of the SRT to perform V and V of a new Digital Rod Position Indication (DRPI) diagnostic system. The DRPI diagnostic system is a digital system that monitors the rod address and rod position data bus voltages of typical DRPI systems in nuclear power plants to detect and diagnose faults. For this application, the SRT was configured to exercise the inputs of the DRPI diagnostic system, and test cases were generated for common system faults such as address and data bit errors, bus voltage, and timing problems. These test cases were applied by the SRT hardware as voltage inputs to the DRPI diagnostic system and the outputs were compared to expected values generated by a model of the DRPI diagnostic system. Included in the paper is a description of the overall design of the SRT including the hardware and software architectures. Methodologies for automating test cases are also described. V and V using the SRT demonstrates the benefits of automated testing and qualification to provide a quantitative assessment of reliability and cost effective implementation of digital I and C in existing and next generation nuclear power plants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Characterization of compositional variation and solidification condition on rapid solidification behavior of beta stabilized titanium

Rapid solidification (RS) is a field of study that has increased in importance due to the rise of additive manufacturing (AM) techniques. The breakdown of equilibrium at the solid-liquid interface results in an inability to use traditional solidification models that have been applied to great success in casting. As such, there is a current need for a high-throughput testing method to screen alloys at rapid solidification conditions without the significant cost for powder atomization. In this work, twin hammer splat quenching (SQ) is utilized and theoretically developed as a technique uniquely placed to test alloy systems for AM applications. Using analytical thermal spray models and numerical heat transfer simulations, the cooling rate and solidification conditions of splat quenching are compared across several parameters to evaluate the key parameters for generating cooling rates relevant to AM process. Once a framework for the cooling rates and relevancy of SQ are developed, SQ was used to process alloy compositions in the Ti5553 alloy space (AMS 7026) to study the formation of detrimental secondary phases occurring upon rapid quenching from the melt. A martensitic phase is found to occur at the lower end of the standard, but the composition listed in the standard were found to be safe. Finally, the rapid solidification behavior of Ti-5553 is evaluated by comparing splat quenching, laser welding, and plasma atomization that offer a wide window of solidification conditions. The segregation and solidification structure are analyzed and compared to models to provide a deeper understanding of the solidification behavior of beta stabilized titanium alloys.

Williamson, Christopher J.↗
DOE OSTI2025

In-situ characterization of a background calibration line for east coast aerial measuring

Report outlines the methodology used to establish a background calibration line for AMS-East operations at Joint Base Andrews. The results provide a general isotopic characterization of the soil within the calibration line at Joint Base Andrews, and the expected terrestrial exposure rate which can be used to aerial measuring.

61 RADIATION PROTECTION AND DOSIMETRY↗