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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 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Magnetic Microcalorimeter Gamma Detectors with Ultra-High Energy Resolution - (Final Report)

The goal of this project was to build a commercially viable magnetic microcalorimeter (MMC) γ-spectrometer with an ultra-high energy resolution <50 eV for accurate non-destructive assay (NDA) of nuclear materials. Improved NDA simplifies and accelerates safeguards analyses in the nuclear fuel cycle and enhances capabilities to detect the diversion of nuclear materials for illicit purposes. High-resolution MMC γ-spectrometers can also improve the accuracy of nuclear data that NDA is based on.

42 ENGINEERING↗
DOE OSTI2023

Massively Parallel Bayesian Model Calibration and Uncertainty Quantification with Applications to Nuclear Fuels and Materials

The U.S. Department of Energy (DOE)’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aims to develop predictive capabilities by applying computational methods to the analysis and design of advanced reactor and fuel cycle systems. This program has been providing engineering-scale support for the development of BISON, a high-fidelity and high-resolution fuel performance tool. Fuel behavior in a nuclear reactor is governed by a complex network of mechanisms interacting with various other physics aspects in the reactor system. Any model developed to represent the fuel behavior will likely be idealized resulting in uncertainties in their predictions compared to the observed data. As such, this report was motivated by the need to identify the sources of uncertainties and quantify and propagate them through the fuel model outputs. Such quantification of uncertainties will establish a level of model trustworthiness, identify approaches to improve the model trustworthiness, and even guide optimal experiment design for maximal information gain. To accomplish the uncertainty quantification for computational models, this report has relied on the Bayesian framework which provides probabilistic treatment of models their inputs and outputs. The current state-of-the-art on performing Bayesian Uncertainty Quantification (UQ) for nuclear engineering models using High Performance Computing (HPC) resources have been reviewed. Implementation of capabilities for massively parallel Bayesian UQ in Multiphysics Object-Oriented Simulation Environment (MOOSE) is discussed. Several verification cases are discussed to verify the accuracy of the quantified uncertainties using the developed computational capabilities in MOOSE. Then, the problem of quantifying the uncertainties in TRI-Structural isOtropic (TRISO) fuel silver release is addressed. For the first time, the uncertainties arising from the TRISO Fission Gas Release (FGR) model due to model inadequacy and experimental noise are quantified. Also, the Bayesian capabilities are applied to the calibration of the MATPRO creep model, a widely used model in several fuel assessment cases. The impact of the prediction uncertainties in the MATPRO model on the fuel cladding behavior as part of the TRIBULATION assessment case (which is an integral effects case) is investigated. This report concludes with a discussion on the future work for the UQ for computational models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Nuclear Criticality Safety Pipeline Course with Hands-On Experimental Training at Lawrence Livermore's Inherently Safe Subcritical Assembly Training Center

The Nuclear Criticality Safety Divisions at Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL) have partnered with Prof. Massimiliano Fratoni of the University of California Berkeley to offer a semester long course on nuclear criticality safety. This course is part of a larger pipeline project among many of the Department of Energy (DOE) laboratories designed to stimulate student interest in the field of criticality safety. The course focuses on teaching the fundamentals of criticality safety, familiarity with national and consensus standards, and preparing criticality safety evaluations. Students also receive hands-on experience with special nuclear material by performing experiments with the Inherently Safe Subcritical Assembly (ISSA) at LLNL. Guest lectures are taught remotely and in-person by criticality safety engineers at LLNL and LANL, giving students an opportunity to interact with professionals in the field. The students complete a semester long project involving developing and writing a criticality safety evaluation. As universities tend to focus heavily on nuclear power and advanced nuclear reactor design, this course gives students a better understanding and perspective of what criticality safety entails. The goal of this pipeline course is to introduce students to criticality safety as another available field for nuclear engineers. It is also a way for criticality safety programs to identify talented students who have the interest and aptitude to work in criticality safety for hire upon graduation. LLNL and LANL have both hosted past students as summer students, participated in student's graduate projects, and hired students as criticality safety engineers. This has provided a unique opportunity for criticality safety programs to spot young talent with better retention outcomes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

24.1.3.1 - Verification of Spent Fuel Inside Dry Storage Casks by Fast/Epi-thermal Neutron Mapping

Development of a spent-fuel verification method based upon positional measurement of fast neutrons emitted from the top surface of the dry storage casks. A proof-of-principle field test on a partially loaded spent-fuel cask will be conducted by measuring the fast neutron flux on the top of the cask as a function of position using a single detector. If a useful verification signature is demonstrated, a multi-detector scanning array will be built and tested in FY2021 to demonstrate a practical rapid verification system in FY2022.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Thermal Neutron Scattering Law (TNSL) Implementation and Testing in FUDGE

Thermal neutron scattering law processing capabilities have been recently implemented in the LLNL code FUDGE (For Updating Data and Generating Evaluations). FUDGE is now capable of producing processed TNSL data for use in Monte Carlo or deterministic transport. To test this new capability, LLNL scientists performed an extensive intercomparison between the Mercury, Ardra, COG and MCNP transport codes. This intercomparison helped probe differences between how TNSL data are handled by the transport codes and by two different processing codes (LLNL’s FUDGE and LANL’s NJOY). This report summarizes recent improvements in FUDGE TNSL capabilities as well as results from the code intercomparison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermal Extraction Modeling within HYBRID: Nominal and Extraction Conditions in Advanced Reactor Systems

This document details continuing development of thermal extraction modeling within the HYBRID repository, the modeling repository of the Integrated Energy Systems program at Idaho National Laboratory. The focus of the thermal extraction work is setting up template energy conversion systems for advanced reactor systems. Based on these nominal conditions, the IES team is developing extraction capability maps that present the system thermal-to-electric efficiency of the remaining system and the mass extraction fraction capability of these systems when extracting steam at various pressures to send steam to applications at a lower given pressure. These curves are mapped on what has become termed “whale chart” due to the appearance of the results. Multiple advanced reactor systems are complete, and it is anticipated that all four of the main advanced reactor types: advanced light-water reactors, high temperature gas reactors, liquid metal fast reactors, and molten salt reactors, will be complete by the end of the fiscal year.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

LLNL International Working Group on Gamma Spectrometry - Preparation of IDB Dataset [Slides]

The spectral information and associated metadata, including source information, detector setup, and shielding configuration is described in human readable, comma-separated text files (csv). Six csv files are associated with each material type and include the following information (where [Isotope] is U, Pu, and MOX): [Isotope]_measurement_metadata Metadata of the measurement configuration, including material composition at time of acquisition (decay corrected from certificate date). Data that is unique to a measurement configuration e.g. detector geometry, source-detector distance, analyzer and other electronics used, attenuating materials are provided in this file. [Isotope]_spectrum_metadata The information that is unique to the spectrum. This includes detector specifications (e.g. date of acquisition, number of channels, gain (keV/channel), energy range, real and live time, full width half-maximum (FWHM) of the 186 or 208 peak for uranium or plutonium measurements, respectively. [Isotope]_source_metadata The source material used for the measurement, this includes the ratio and uncertainties of isotopes present and processing information of the material. In addition, the certificate information for the source material is listed. [Isotope]_spectrum_counts_metadata The counts per channel for each spectrum. This data is linked via the UID.spectrum index to the measurement_metadata file. [Isotope]_spectrum_checksum_metadata The MD5 checksum for the original spectrum file. This data is linked via the UID.spectrum index to the measurement_metadata file. This information is provided to prevent duplication of entries in the database. [Isotope]_provenance Details on the csv data provider, including laboratory and contract information.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Characterization of Impure Plutonium Oxide Neutron Calibration Standards through Evaluation by Neutron Multiplicity Equations

Plutonium oxide calibration standards STD4C, STD6C and STD8C have been historically used, and are currently the only MT52 standards in use, for calibration and measurement control program of Passive/Active Neutron Coincidence Counter known as PAN1 at PF-4. In this paper, previously measured high-statistics data are used in conjunction with neutron multiplicity equations to determine multiplication M and α parameter of these standards as the explicit knowledge of these parameters may allow quantification of precision and bias of accountability measurements of different material types and matrices of items.

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DOE OSTI2024

Hazards and Probabilistic Risk Assessments of Advanced Nuclear Reactors Coupled with Industrial Facilities

This report provides a roadmap and tool kit for site specific risk assessments across a broad range of industrial customers co-located with advanced nuclear power plants (ANPP) that are not currently built and operating in the U.S. This report builds upon the body of work sponsored by the Department of Energy (DOE) Integrated Energy Systems Pathway that has produced industrial requirements studies and techno-economic assessments on the topics of feasibility of ANPP supported industrial processes. This report also leverages the DOE Light Water Reactor Sustainability (LWRS) program that has presented hazards assessment and generic probabilistic risk assessments (PRAs) for the addition of a heat extraction system (HES) to light-water reactors (LWRs) co-located with hydrogen production facilities. Many of the hazard assessments and risk assessments performed for the LWRS report are agnostic to whether the nuclear reactor is an ANPP or were adapted to the ANPP focus. The report performs hazards assessments to include industrial facilities: an oil refinery, a methanol plant, a synthetic fuel (synfuel) plant, the production of synthetic gas (syngas) as part of the methanol and synfuel plants, wood pulp and paper mills, and hydrogen production. Hydrogen production facilities are assessed in depth through prior reports in the LWRS program and the results are leveraged in this report. All these facilities are specified through industrial process and requirements research performed by national laboratories, universities, and interaction with industry. Many of the processes used in this report are pre-conceptual designs to use for decarbonization of the current technology facilities. A process of failure modes and effects analysis (what can go wrong) and accidentology (what has historically gone wrong) was used to determine the hazards presented to the nuclear power plant by the addition of the HES and the industrial customer. Chemical properties of feedstocks and products are summarized as part of the hazards assessment. Example analysis procedures are provided for each of the hazard types identified. These deterministic analyses can be used to assess adherence to licensing criteria. They can also be used to meet other safety goals like protection of the public, workers, or industrial facility equipment. A modular high temperature gas-cooled reactor (MHTGR) PRA only existing on paper was modeled and verified in modern PRA software. This will provide a tool for representative ANPP probabilistic analyses for future research.

10 SYNTHETIC FUELS↗
NASA NTRS1985

Synthetic Organic Materials in Nuclear Powerplants

Report aids plant designers and qualification engineers in ensuring that organic materials in nuclear powerplants will perform satisfactorily in such safety-related equipment as insulation on motor windings, pump diaphragms, motor and pump lubricants, and pump seals and gaskets. Report provides information for service that may include both mild and harsh nuclear environments.

Bouquet, F. L.↗
DOE OSTI2022

An Overview in the Development of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

To enhance nuclear nonproliferation stewardship, Idaho National Laboratory is building the Beartooth nuclear fuel cycle processing test bed. The Beartooth test bed will allow researchers the opportunity to study nuclear fuel processing operations including the use of centrifugal contactors in solvent extraction processes. The test bed is designed to support data collection and machine learning to monitor process operations. As part of this project, researchers will study data collected from a host of sensors that have not been typically used to monitor solvent extraction processes such as vibration, acoustic, colormetric, and thermal measurement data. The goal of this research is to employ machine learning and data analytics to study the confluence of signals collected from both traditionally and non-traditionally used sensors to provide operator process awareness and discover process anomalies. An overview of planned sensors and experiments that will focus on signal discovery will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

An Overview in the Development of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

To enhance nuclear nonproliferation stewardship, Idaho National Laboratory is building the Beartooth nuclear fuel cycle processing test bed. The Beartooth test bed will allow researchers the opportunity to study nuclear fuel processing operations including the use of centrifugal contactors in solvent extraction processes. The test bed is designed to support data collection and machine learning to monitor process operations. As part of this project, researchers will study data collected from a host of sensors that have not been typically used to monitor solvent extraction processes such as vibration, acoustic, colormetric, and thermal measurement data. The goal of this research is to employ machine learning and data analytics to study the confluence of signals collected from both traditionally and non-traditionally used sensors to provide operator process awareness and discover process anomalies. An overview of planned sensors and experiments that will focus on signal discovery will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

ATR Fuel Management

The sixth Core Internals Changeout (CIC) of ATR was completed on March 21, 2022, and Nuclear Testing (NT) commenced shortly thereafter. In support of NT, 235U flux wires were attached to flux wands using electrical tape and installed in water channels 2, 6, 11, 15 and 19 in each of the 40 ATR Mark VII fuel elements that comprise an ATR core to perform low power flux measurements during ATR Cycle 170CIC NT-3. The flux wands were fabricated from Poly (Methyl Methacrylate), or PMMA. Irradiated cobalt capsules were loaded in the ATR Northeast Irradiation Housing Assembly (NEIHA), the East Flux Trap (EFT), South Flux Trap (SFT), inboard A-positions and various H-positions as a 170CIC NT-3 core reactivity hold-down mechanism. The PMMA flux wands received significant gamma dose from the irradiated cobalt during NT-3 which caused discoloration, embrittlement and swelling of the flux wands. The removal of the degraded flux wands from the 0.078 in. wide by 50 in. long fuel element coolant channels has been a unique challenge. This presentation summarizes the cleaning techniques used to remove the flux wand debris from the elements, an aggressive PMMA irradiation experiment performed concomitantly with initial cleaning efforts, and planned inspection and qualification techniques that will be applied in the future to potentially requalify the impacted elements for re-use.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2018

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Taylor, Brian↗
NASA NTRS2012

Nuclear Thermal Propulsion (NTP): A Proven Growth Technology for Human NEO/Mars Exploration Missions

The nuclear thermal rocket (NTR) represents the next "evolutionary step" in high performance rocket propulsion. Unlike conventional chemical rockets that produce their energy through combustion, the NTR derives its energy from fission of Uranium-235 atoms contained within fuel elements that comprise the engine s reactor core. Using an "expander" cycle for turbopump drive power, hydrogen propellant is raised to a high pressure and pumped through coolant channels in the fuel elements where it is superheated then expanded out a supersonic nozzle to generate high thrust. By using hydrogen for both the reactor coolant and propellant, the NTR can achieve specific impulse (Isp) values of ~900 seconds (s) or more - twice that of today s best chemical rockets. From 1955 - 1972, twenty rocket reactors were designed, built and ground tested in the Rover and NERVA (Nuclear Engine for Rocket Vehicle Applications) programs. These programs demonstrated: (1) high temperature carbide-based nuclear fuels; (2) a wide range of thrust levels; (3) sustained engine operation; (4) accumulated lifetime at full power; and (5) restart capability - all the requirements needed for a human Mars mission. Ceramic metal "cermet" fuel was pursued as well, as a backup option. The NTR also has significant "evolution and growth" capability. Configured as a "bimodal" system, it can generate its own electrical power to support spacecraft operational needs. Adding an oxygen "afterburner" nozzle introduces a variable thrust and Isp capability and allows bipropellant operation. In NASA s recent Mars Design Reference Architecture (DRA) 5.0 study, the NTR was selected as the preferred propulsion option because of its proven technology, higher performance, lower launch mass, versatile vehicle design, simple assembly, and growth potential. In contrast to other advanced propulsion options, no large technology scale-ups are required for NTP either. In fact, the smallest engine tested during the Rover program - the 25,000 lbf (25 klbf) "Pewee" engine is sufficient when used in a clustered engine arrangement. The "Copernicus" crewed spacecraft design developed in DRA 5.0 has significant capability and a human exploration strategy is outlined here that uses Copernicus and its key components for precursor near Earth object (NEO) and Mars orbital missions prior to a Mars landing mission. The paper also discusses NASA s current activities and future plans for NTP development that include system-level Technology Demonstrations - specifically ground testing a small, scalable NTR by 2020, with a flight test shortly thereafter.

Borowski, Stanley K.↗
DOE OSTI2021

Comparing Critical Mass of Homogeneous and Heterogeneous Systems

The purpose of this project is to recreate a critical mass study comparing homogeneous and heterogeneous systems using a conservative modeling approach to expand upon historical experimental data. This project will help Nuclear Criticality Safety (NCS) engineers determine if modeling a heterogeneous system comprised of uranium and water as homogeneously mixed at a comparable density is a conservative method.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2021

Sequential Bayesian inference of transition rates in the hidden Markov model for multi-state system degradation

The more easily available system performance data and advances in data analytics have provided us with opportunities to optimize maintenance programs for engineered systems, for example nuclear power plants. One key task in maintenance optimization is to obtain an accurate model for system degradation. Here, in this research, we propose a Bayesian method to address this problem. Noting that systems usually exhibit multiple states and that the actual state of a system usually is not directly observable, in the method we first model the system degradation process and the observation process based on a hidden Markov model. Then we develop a sequential Bayesian inference algorithm based on importance sampling and the forward algorithm to infer the posterior distributions of the transition rates in the hidden Markov model based on available observations. The proposed Bayesian method allows us to take advantage of evidence from multiple sources, and also allows us to perform Bayesian inference sequentially, without the need to use the entire history of observations every time new observations are collected. We demonstrate the proposed method using both synthetic data for a nuclear power plant feedwater pump and realistic data for a nuclear power plant chemistry analytical device.

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