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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Thermal Analysis Capability of UNF-ST&DARDS

In the United States, approximately 2,500 casks are loaded with commercial spent nuclear fuel (SNF) that has transitioned from wet storage (spent fuel pools) to dry storage. The number of loaded dry storage casks is increasing by approximately 200 each year. Over time, cask designs have evolved to enhance safety and to accommodate more fuel and higher heat loads. Also, higher burnup fuel is being transitioned into dry storage. The SNF is being stored in dry casks for longer times than specified in the original certification period. Several degradation mechanisms related to fuel assemblies and canisters are affected by temperature. For the cladding, temperature-dependent phenomena include creep and annealing, hydride reorientation and embrittlement, and the ductile-to-brittle transition. Temperature can also influence phenomena that affect the long-term integrity of the storage system, including deliquescence, corrosion, and stress-corrosion cracking. Therefore, accurate determination of the temperatures of various components is needed to evaluate potential safety-related issues during transportation after extended storage and to ensure SNF retrievability. The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is being developed for the US Department of Energy Office of Nuclear Energy to streamline analyses for the waste management system [1]. Lastly, the thermal analysis capability within UNF-ST&DARDS and example results are discussed herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Observing terrestrial ecosystems and the carbon cycle from space

Modeled terrestrial ecosystem and carbon cycle feedbacks contribute substantial uncertainty to projections of future climate. The limitations of current observing networks contribute to this uncertainty. Here we present a current climatology of global model predictions and observations for photosynthesis, biomass, plant diversity and plant functional diversity. Carbon cycle tipping points occur in terrestrial regions where fluxes or stocks are largest, and where biological variability is highest, the tropics and Arctic/Boreal zones. Global observations are predominately in the mid-latitudes and are sparse in high and low latitude ecosystems. Observing and forecasting ecosystem change requires sustained observations of sufficient density in time and space in critical regions. Using data and theory available now, we can develop a strategy to detect and forecast terrestrial carbon cycle-climate interactions, by combining in situ and remote techniques.

carbon cycle↗
DOE OSTI2012

Conceptual Ideas for New Nondestructive UF 6 Cylinder Assay Techniques

Nondestructive assay (NDA) measurements of uranium cylinders play an important role in helping the International Atomic Energy Agency (IAEA) safeguard uranium enrichment plants. Traditionally, these measurements have consisted of a scale or load cell to determine the mass of UF{sub 6} in the cylinder combined with a gamma-ray measurement of the 186 keV peak from 235 U to determine enrichment. More recently, Los Alamos National Laboratory (LANL) and Pacific Northwest National Laboratory (PNNL) have developed systems that exploit the passive neutron signal from UF 6 to determine uranium mass and/or enrichment. These include the Uranium Cylinder Assay System (UCAS), the Passive Neutron Enrichment Meter (PNEM), and the Hybrid Enrichment Verification Array (HEVA). The purpose of this report is to provide the IAEA with new ideas on technologies that may or may not be under active development but could be useful for UF 6 cylinder assay.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Review of Destructive Assay Methods for Nuclear Materials Characterization from the Three Mile Island (TMI) Fuel Debris

This report provides a summary of the literature review that was performed and based on previous work performed at the Idaho National Laboratory studying the Three Mile Island 2 (TMI-2) nuclear reactor accident, specifically the melted fuel debris. The purpose of the literature review was to document prior published work that supports the feasibility of the analytical techniques that were developed to provide quantitative results of the make-up of the fuel and reactor component debris located inside and outside the containment. The quantitative analysis provides a technique to perform nuclear fuel accountancy measurements

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Final Report For Project TCF-18-15778

This report documents work performed under the Technology Commercialization Fund (TCF-18-15778, Used (spent) nuclear fuel management and analysis tool) award provided by the US Department of Energy (DOE) Office of Technology Transitions (OTT). This report describes various Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) enhancements to advance the technology readiness level (TRL). The principal work performed under TCF is the integration of the initial dry storage loading optimization algorithm into UNF-ST&DARDS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AFIP6-MkII and RERTR-12 Porosity Data Collection and Analysis for Modeling and Simulation

Gathering data for the improvement of nuclear fuel modeling and simulation efforts is the primary driver for this work. Mechanistic models allow for a better understanding of the material on a micro- and macrostructural level while saving time and money over traditional experiment efforts. Historically, summarized data and correlations are the inputs for empirical material models and model validation. When improving these models for nuclear fuels with experimental results, there is a lack of reliable data readily available. Experiments - RERTR-12 and AFIP6-MkII - were conducted to understand the irradiation behavior of metallic U-10Mo monolithic fuels for use in extreme reactor environments such as research reactors like the Advanced Test Reactor (ATR) or the High Flux Isotope Reactor (HFIR). Microstructural characteristics of fission gas pores (FGP) in each experiment are collected using an automated image analysis technique developed at the University of Florida and presented here. A series of statistical tests are performed to explore the reliability of the results, as well as understand where the data is lacking and what future data collection is necessary to provide sufficient information to assist modeling efforts. The focus is on the porosity, pore size, and eccentricity of FGPs formed during irradiation in three AFIP6-MkII samples and one RERTR-12 sample. From the analysis, it is clear there are substantial impacts of fission density on the pore structure, but there also exist also underlying connections between each sample and the behavior observed in the pores. Further analyses of the pre- and post-irradiation microstructure are needed to improve the understanding of these connections. An early method for microstructural data analysis is presented within and is currently being expanded to include other microstructure data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Demonstration of the On-the-Fly Shielding Analysis Method: Spent Fuel and Waste Disposition

This report documents work performed supporting the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management activities under work breakdown structure element 1.08.02.04.01, “Data and Tools Development, Validation, and Maintenance.” In particular, this report fulfills milestone M3SF-21OR020401016, “Implement on-the-fly dose analysis methodology in UNF-ST&DARDS” within work package SF-21OR02040101, “Commercial SNF Characterization - ORNL.” The Used Nuclear Fuel - Storage, Transportation & Disposal Analysis Resource and Data System (UNFST& DARDS) enables automated dose rate calculations for spent nuclear fuel (SNF) transportation packages and storage casks using a Monte Carlo radiation transport code. The explicit method uses a detailed model of the SNF system and its contents. Therefore, a dose rate calculation is required for each as-loaded transportation package or storage cask because the SNF assemblies within a canister typically have unique irradiation characteristics. An alternate method, referred to as the “on-the-fly” shielding analysis method, has been proposed that requires only a set of Monte Carlo dose rate calculations for each transportation packaging/storage cask design. The results of the Monte Carlo dose rate calculations are independent of the SNF assembly irradiation and decay characteristics. The dose rate values may then be combined with the radiation source strength of the SNF assemblies associated with a particular transportation packaging/storage cask design to determine actual dose rates. This report presents on-the-fly dose rate calculations for a representative SNF storage cask and verification of the on-the-fly dose rate calculation results by comparison with reference dose rate calculations using the explicit Monte Carlo dose rate calculation. The on-the-fly shielding analysis method was implemented in UNF-ST&DARDS. A Python program was developed to process the MAVRIC dose rate results obtained by source particle type, energy group, and fuel geometry region. A Python processor created binary files, which were saved as a special UNF-ST&DARDS library for on-the- fly shielding analyses. UNF-ST&DARDS uses the precalculated on-the-fly binary libraries generated by the Python data processor and directly executes the Python code for on-the-fly dose analysis. This Python code unzips the pre-generated binary files mentioned above, reads the data, and combines them with user-specified sources for dose and uncertainty calculations. The Python programs were verified using Excel calculations and by comparison with the values obtained with the MAVRIC post-processing utilities applied to the 3dmap files. This method can currently be used to determine dose rates for as-loaded HI-STORM FW storage casks. The UNF-ST&DARDS analysis wizard for on-the-fly shielding analysis is described in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2019

Progress Toward the Critical Design of the Superconducting Rotor for NASA's 1.4 MW High Efficiency Electric Machine

NASA Glenn Research Center is developing a 1.4 MW high-efficiency electric machine for future electrified aircraft to reduce energy consumption, emissions, and noise. This wound-field, synchronous machine employs a self-cooled, superconducting rotor to achieve excellent specific power and efficiency. This paper presents the progress made over the past year toward a critical design of the superconducting rotor, including refined analysis and risk reduction testing on prototype superconducting coils. A 3D finite element model of the rotor is developed to simulate the combined loading imposed by the initial cool down of the rotor from room temperature to cryogenic temperature and the subsequent rotation of the rotor up to the design speed. The model is evaluated to better predict the stress state in the critical components and the reduction in physical gap between the stator and rotor due to the resulting radial deflections. This model includes an enhanced model of the superconducting coil that considers the coil as an assembly of four individual coil layers. An improved coil fabrication process is described in detail. Two sub-scale test coils are fabricated and then tested under repeated thermal cycling between room temperature and liquid nitrogen temperature (77 K). It is demonstrated that the proposed coil fabrication procedure can produce no-insulation, high temperature superconducting coils that can survive several thermal cycles without apparent degradation.

Scheidler, Justin↗
DOE OSTI2011

First Results from the Seven Percent Critical Experiment.

The Seven Percent Critical Experiment (7uPCX) was designed to provide benchmark criticality and reactor physics data for water-moderated pin-fueled nuclear reactor cores. The enrichment of the fuel was chosen to explore the enrichment range above the current 5% ceiling for US commercial PWRs. The experiment is part of the Nuclear Energy Research Initiative (NERI) Project 01-124 titled “Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel”. The NERI project is a collaboration between AREVA Federal Services, LLC; the University of Florida; Oak Ridge National Laboratory; and Sandia National Laboratories (SNL). The experiments are currently supported by the National Nuclear Security Administration Nuclear Criticality Safety Program. This paper presents the first set of results from the 7uPCX arrays performed at SNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Results from the First Set of Criticals In the Seven Percent Critical Experiment [Slides]

This presentation discusses the recent Sandia critical experiments, specifically the Seven Percent Critical Experiment (7uPCX) which is a Nuclear Energy Research Initiative (NERI) project. It also discusses why 7uPCX was used, how 7uPCX is operated and presents some 7uPCX results. The presentation concludes by discussing future plans for the critical experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Alternative methods for degradation studies by alpha radiolysis: tributyl phosphate and CMPO

Solvent extraction separation processes used in the recycling of used nuclear fuel are susceptible to radiolytic damage from radioactive isotopes present in used fuel. Studying the respective effects on matter of both low linear energy transfer (LET) radiation such as gamma radiation and high LET such as alpha radiation will allow for accurate prediction and modeling of process performance losses with respect to dose. The effects of gamma radiation on solvent extraction ligands have been more extensively studied than the effects of alpha radiation due to the inherent difficulty in producing a sufficient and confluent dose of alpha particles within a sample without leaving the sample contaminated with long lived radioactive isotopes. We have developed a method for studying the effects of high LET radiation in situ via {sup 10}B activation and the high LET particles that result from the {sup 10}B(n,a){sup 7}Li reaction which follows. In this study we applied this method to organic solutions of tributyl phosphate (TBP) and CMPO (compound octylphenyl-N, N-diisobutyl-carbamoyl methyl phosphine oxide) representing the PUREX and TRUEX processes respectively. Rates of degradation of TBP and CMPO and their respective degradation products in the presence of both high and low LET radiation are presented and compared to values reported in the literature. Preliminary data appears to show decreased degradation of CMPO in the presence of an aqueous acidic phase, which agrees with other studies performed on TBP solutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Employing Whisper in selecting benchmarks for traditional validation methods

The Nuclear Criticality Safety Program at Sandia National Laboratories (SNL) uses a collection of benchmarks developed from the International Handbook of Evaluated Criticality Safety Benchmark Experiments for validation of MCNP. In early 2016, an effort began to expand the benchmark collection to cover a broader range of materials and neutron spectra. The program undertook the task of updating current benchmarks to reflect best practices from Los Alamos National Laboratory (LANL) for MCNP calculations. Spring of 2016, LANL shared the Whisper code with SNL in a collaborated effort to investigate application of the code to activities at SNL. Whisper is a statistical package developed to support nuclear criticality safety validation efforts. It measures how well benchmarks correlate to a system or process and calculates a USL. Whisper has begun to develop into an asset for the SNL Nuclear Criticality Safety team to determine and reassure benchmarks selected for validation truly correlate process parameters (i.e., fall within an acceptable Area of Applicability). SNL's use of traditional statistical methods, with the abilities of Whisper to identify correlating benchmarks, has facilitated in ensuring confidence in validation and USL determination. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRSDate not supplied

Learning From Failure: Boosting Cycling Endurance of Optical Phase Change Materials

Chalcogenide phase change materials (PCMs) are a unique class of compounds whose switchable optical and electronic properties have fueled an explosion of emerging applications in microelectronics and microphotonics. Key to any application is the ability of PCMs to reliably switch between crystalline and amorphous states over a large number of cycles. While this issue has been extensively studied in the case of microelectronic memories, current PCM-based optical devices suffer from much inferior endurance. To understand the failure mechanisms limiting endurance of PCMs specifically in microphotonic devices, we have developed an on-chip resistive micro-heater platform and an automatic multi-modal characterization system to analyze cycling performance of optical PCMs. Reversible switching of large-area PCM devices over 50,000 cycles was demonstrated.

Optical phase change material↗