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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Examining supply chain resilience for the intermodal shipment of spent nuclear fuel and high level radioactive materials-16309

The U.S. Department of Energy (DOE) has a significant programmatic interest in the safe and secure routing and transportation of Spent Nuclear Fuel (SNF) and High Level Waste (HLW) in the United States, including shipments entering the country from locations outside U.S borders. In any shipment of SNF/HLW, there are multiple chains; a jurisdictional chain as the material moves between jurisdictions (state, federal, tribal, administrative), a physical supply chain (which mode), as well as a custody chain (which stakeholder is in charge/possession) of the materials being transported. Given these interconnected networks, there lies vulnerabilities, whether in lack of communication between interested stakeholders or physical vulnerabilities such as interdiction. By identifying key links and nodes as well as administrative weaknesses, decisions can be made to harden the physical network and improve communication between stakeholders. This paper examines the parallel chains of oversight and custody as well as the chain of stakeholder interests for the shipments of SNF/HLW and the potential impacts on systemic resiliency. Here, by using the Crystal River shutdown location as well as a hypothetical international shipment brought into the United States, this paper illustrates the parallel chains and maps them out visually.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

43 PARTICLE ACCELERATORS↗
DOE OSTI2018

Novel capabilities for examining and summarizing hierarchical and geographic resiliency profiles

The geospatial research and development team in the National and Homeland Security Division at Idaho National Laboratory was tasked with providing tools to derive insight from the substantial amount of data currently available - and continuously being produced - associated with the critical infrastructure of the US. This effort is in support of the Department of Homeland Security, whose mission includes the protection of this infrastructure and the enhancement of its resilience to hazards, both natural and human. We present geovisual-analytics-based approaches for analysis of vulnerabilities and resilience of critical infrastructure, designed so that decision makers, analysts, and infrastructure owners and managers can manage risk, prepare for hazards, and direct resources before and after an incident that might result in an interruption in service. Our designs are based on iterative discussions with DHS leadership and analysts, who in turn will use these tools to explore and communicate data in partnership with utility providers, law enforcement, and emergency response and recovery organizations, among others. In most cases these partners desire summaries of large amounts of data, but increasingly, our users seek the additional capability of focusing on, for example, a specific infrastructure sector, a particular geographic region, or time period, or of examining data in a variety of generalization or aggregation levels. These needs align well with tenets of in-formation-visualization design; in this paper, selected applications among those that we have designed are described and positioned within geovisualization, geovisual analytical, and information visualization frameworks.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Irradiated AGR-2 Compact 3-1-1 Examination Plan

This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 3-1-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan and guidance in the ORNL PIE statement of work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Irradiated AGR 2 Compact 4-1-2 Examination Plan

This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 4-1-2 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the Statement of Work for PIE and Safety Testing of Pebble Bed Modular Reactor Limited (PBMR) Fuel at ORNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Irradiated AGR-2 Compact 6-4-2 Examination Plan

This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 6-4-2 that was taken from the Advanced Gas Reactor (AGR) experiment, AGR-2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan and guidance in the ORNL PIE statement of work

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Irradiated AGR 2 Compact 6 2 2 Examination Plan

This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 6-2-2 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan and guidance in the ORNL PIE statement of work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Irradiated AGR-2 Compacts 5-1-1, 5-1-2, and 5-3-1 Examination Plan

This plan describes post-irradiation examination (PIE) activities (simultaneous three-compact temperature transient safety testing and optional post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compacts 5-1-1, 5-1-2, and 5-3-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan and guidance in the ORNL PIE statement of work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Irradiated AGR-2 Compact 6-4-3 Examination Plan

This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 6-4-3 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan0F and guidance in the ORNL PIE statement of work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Irradiated AGR-2 Compact 5-2-1 Examination Plan

This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 5-2-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan0F and guidance in the ORNL PIE statement of work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-3/4 Compact 1-4 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 1-4 taken from the Advanced Gas Reactor (AGR) experiment, AGR-3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Irradiated AGR 2 Compact 4-3-1 Examination Plan

This plan describes post-irradiation examination activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory on irradiated Compact 4-3-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the Statement of Work for PIE and Safety Testing of Pebble Bed Modular Reactor Limited Fuel at ORNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Examining Perceptual Luminance Uniformity of Simulated Luminaire Patterns

Luminaire luminance uniformity is an important aspect that can affect perceived lighting quality, discomfort glare, and efficacy. While several metrics have been proposed to characterize luminance uniformity, previous studies have shown that current metrics such as Max:Min or Avg:Min luminance ratios can be insensitive to important differences in luminance gradient that may affect perceived uniformity. In an attempt to resolve this issue, previous studies incorporated a contrast sensitivity function for the human eye based on spatial frequency, such as in the UHVS metric; however, this metric has not been comprehensively studied in relation to perceived uniformity ratings. The study presented in this paper aimed to examine the relationship between the UHVS metric and perceived uniformity ratings. Specifically, the study used a web-based questionnaire that presented simulated luminance patterns and asked participants to assess uniformity using a two-alternative forced-choice procedure. Responses were collected and analyzed from 94 participants. The results showed a significant correlation and relationship between the UHVS metric and perceived uniformity. However, comparisons between patterns that had similar UHVS and between patterns that had larger differences in UHVS did not yield consistent results. These results suggest that the UHVS metric might be used for general guidance but may warrant further studies to better understand its sensitivity and improve its alignment with perceived uniformity ratings.

Abboushi, Belal K.↗
DOE OSTI2021

Examining the Use of the Technology Acceptance Model for Adoption of Advanced Digital Technologies in Nuclear Power Plants

The United States nuclear industry is being economically challenged and needs wide-scale adoption of enabling technology like advanced automation to remain viable. To enable wide-scale adoption of advanced technology, addressing the human-technology integration challenges and factors that influence technology acceptance (i.e., ‘buy in’) is pertinent. This work examines the Technology Acceptance Model as a framework for characterizing the factors influencing technology adoption in the nuclear industry. An outcome of this work is to support the development of future tools such as a technology acceptance survey that can be used to assess technology acceptance throughout the lifespan of a modernization program to provide a systematic way of ensuring stakeholder and operations ‘buy in.’

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

AGR-3/4 Compact 10-4 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 10 4 taken from the Advanced Gas Reactor (AGR) experiment, AGR 3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Examination Of IASCC In 304 Stainless Steel Core Shroud From A Commercial Boiling Water Reactor (BWR)

The 304 stainless steel core shroud and the accompanying weld material from a commercial boiling-water reactor (BWR) experienced an estimated dose of ~3.5 dpa neutron irradiation leading to irradiation-assisted stress corrosion cracking (IASCC). The resulting IASCC crack tips are examined with high-resolution analysis techniques, including scanning electron microscopy (SEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT), with select examples of directly correlated STEM/APT observations. Analytical observations reveal the crack morphology, composition, microstructure, and any associated corrosion-induced or radiation-induced segregation in the local metal grain boundaries ahead of the propagating SCC cracks. The observed oxide composition and structure were in-line with expectations, consisting of a Cr-rich inner oxide and a Fe-rich outer oxide, with the leading crack tip oxide being Cr-rich. In addition, the leading grain boundary exhibited typical radiation-induced segregation signatures that were largely unperturbed by the accompanying corrosion. Finally, the quantitative capabilities and limitations of STEM and APT in the context of corrosion fronts and grain boundary chemistry from the directly correlated measurements are discussed.

Olszta, Matthew J.↗
DOE OSTI2020

AGR-2 Compact 6-4-1 Post-Irradiation Examination Results

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2017). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗