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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A regulatory perspective on the response function method for determining allowable content of a radioactive materials transportation package or storage cask - 15023

In order to simplify shielding design, there has been an increasingly use of a so-called dose rate response function method for determining the allowable contents in radioactive materials (including spent fuel) transportation package and spent fuel storage cask designs. In essence, this approach uses a pre-calculated dose rate at a given location outside the storage cask or transportation package for a given source inside the cask. It assumes that there is a one-to-one relationship between a source inside the cask and the dose rate outside the cask, i.e., a dose rate response function of a source. Thus, the allowable content is obtained by simply dividing the targeted dose rate by the response function. This method converts the complex particle transport problem into a simple arithmetic calculation. If used correctly and prudently, this approach can greatly simplify the shielding design of casks. However, the use of this method must demonstrate the validity of the assumption, that is, the nuclear physics property of the media through which the particle transverses must be similar to that used in computing the response function. Because of the assumptions and approximations, either implicitly or explicitly, made in the models for dose rate response functions, to assure safety the users should also consider the errors in the models and the bias and uncertainties associated with the code used to calculate the response function. This paper reviewed the advantages and some of the major potential pitfalls of this method from the regulatory perspective. (authors)

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

Validation and Independent Uncertainty Analysis of the MIX-SOL-THERM-003 ICSBEP Benchmark

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was started in 1992 by the United States Department of Energy and later in 1995 became an international project with contributions from 22 countries. The project is now organized by the OECD (Organisation for Economic Co-operation and Development) Nuclear Energy Agency (NEA). In its most recent iteration, the ICSBEP handbook contains over five thousand evaluations of critical, near-critical, and subcritical experiments conducted in facilities all around the world. These benchmarks serve as valuable information for criticality safety engineers who can use them to validate calculation techniques and establish minimum subcritical margins for operations with fissionable materials. The benchmarks in the handbook are categorized by their fissile material composition, material form (oxide, solution, or metal), and fission energy spectra. This is especially useful for those looking for benchmarks similar to a system they are working on to compare methods and identify trends. The ICSBEP Handbook Uncertainty Guide is document outlining recommended practices and methods for determining uncertainties in these benchmarks. Quantifying these uncertainties thoroughly is crucial as it allows a higher degree of confidence that data used from them is valid and relevant. The guide stresses the importance of a thorough and well documented uncertainty analysis when evaluating an experiment. All measured values of a system, whether they be dimensions or material compositions, have a certain amount of uncertainty associated with them and can be analyzed one by one to determine their effects on the system. Many evaluated benchmarks in the handbook present this in detail, however some do not, mostly earlier evaluations performed in the 1990’s and early 2000’s. Recently at Los Alamos National Laboratory (LANL), the Nuclear Criticality Safety Division (NCSD) of LANL has been validating MCNP6.2 ® input files of criticality benchmarks for use by Whisper, a criticality safety code developed at LANL. This effort is also part of the OECD NEA Working Party on International Nuclear Data Evaluation Co-operation (WPEC) Subgroup 45, also known as Validation of Nuclear Data Libraries (VaNDaL). The goal of VaNDaL is to compile a set of validated simulation inputs for use in validating nuclear data and simulation codes. As part of these efforts, one of the benchmarks reviewed was the MIX SOL-THERM-003 ICSBEP benchmark. This paper provides an independent uncertainty analysis of this benchmark experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m 2 . Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

Erosion–deposition↗
DOE OSTI2021

Review of WEC-Sim Development and Applications

WEC-Sim (Wave Energy Converter Simulator) is an open-source code for simulating wave energy converters that has been actively developed and applied to simulate a wide variety of device archetypes, and has become a popular tool since its release. This paper reviews the development efforts and usage of WEC-Sim. The publications considered in this study have been broken down into six topic areas, namely, feature development, experimental validation, device modelling, control modelling, powertake-off (PTO) and grid modelling, and novel applications, which includes some non-wave energy applications. This review paper also recognizes the contributions of academic researchers and technology developers from around the world toward the broader WEC-Sim development effort. The growing number of external applications of WEC-Sim demonstrates a broader acceptance of the open-source code, and the ways WEC-Sim has been used in certain topic areas also highlight potential future development needs.

applications↗
DOE OSTI2024

Analysis of the ASME Code Rules for Subsection III-5-HHB (Composite Materials) for Current HTR Design Requirements

This document includes the critical analysis review of the American Society of Mechanical Engineers (ASME) Section III Division 5 Subsection HH Subpart B (HHB), including Mandatory Appendices, that was published in 2023. In the context of this document, reference to “the code” is specific to this subsection unless otherwise specified. A specific composites task group within the ASME Nonmetallic Design and Materials Working Group, with the support of external experts, was established to perform a gap analysis review. The significant findings are summarized here. The committee response with suggested action items are detailed in the body of the report.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Review of WEC-Sim development and applications

WEC-Sim (Wave Energy Converter Simulator) is an open-source code for simulating wave energy converters, which has been actively developed and applied to simulate a wide variety of device archetypes, and has become a popular tool since its release. This paper reviewed the development efforts and the usage of WEC-Sim. The publications considered in this study have been broken down into six topic areas, namely feature development, experimental validation, device modeling, control modeling, PTO and grid modeling, and novel applications, which even includes some non-wave energy applications. This review paper has also attempted to recognize the contributions of the broader WEC-Sim development effort, meaning not only the internal WEC-Sim development team but also the external efforts from the academia researchers and technology developers around the world. The growing trend of external applications of WEC-Sim has demonstrated the broader acceptance of the open-source code, and how WEC-Sim has been used in a certain topic area also highlights the potential future development needs.

16 TIDAL AND WAVE POWER↗
DOE OSTI2011

NA-42 TI Shared Software Component Library FY2011 Final Report

The NA-42 TI program initiated an effort in FY2010 to standardize its software development efforts with the long term goal of migrating toward a software management approach that will allow for the sharing and reuse of code developed within the TI program, improve integration, ensure a level of software documentation, and reduce development costs. The Pacific Northwest National Laboratory (PNNL) has been tasked with two activities that support this mission. PNNL has been tasked with the identification, selection, and implementation of a Shared Software Component Library. The intent of the library is to provide a common repository that is accessible by all authorized NA-42 software development teams. The repository facilitates software reuse through a searchable and easy to use web based interface. As software is submitted to the repository, the component registration process captures meta-data and provides version control for compiled libraries, documentation, and source code. This meta-data is then available for retrieval and review as part of library search results. In FY2010, PNNL and staff from the Remote Sensing Laboratory (RSL) teamed up to develop a software application with the goal of replacing the aging Aerial Measuring System (AMS). The application under development includes an Advanced Visualization and Integration of Data (AVID) framework and associated AMS modules. Throughout development, PNNL and RSL have utilized a common AMS code repository for collaborative code development. The AMS repository is hosted by PNNL, is restricted to the project development team, is accessed via two different geographic locations and continues to be used. The knowledge gained from the collaboration and hosting of this repository in conjunction with PNNL software development and systems engineering capabilities were used in the selection of a package to be used in the implementation of the software component library on behalf of NA-42 TI. The second task managed by PNNL is the development and continued maintenance of the NA-42 TI Software Development Questionnaire. This questionnaire is intended to help software development teams working under NA-42 TI in documenting their development activities. When sufficiently completed, the questionnaire illustrates that the software development activities recorded incorporate significant aspects of the software engineering lifecycle. The questionnaire template is updated as comments are received from NA-42 and/or its development teams and revised versions distributed to those using the questionnaire. PNNL also maintains a list of questionnaire recipients. The blank questionnaire template, the AVID and AMS software being developed, and the completed AVID AMS specific questionnaire are being used as the initial content to be established in the TI Component Library. This report summarizes the approach taken to identify requirements, search for and evaluate technologies, and the approach taken for installation of the software needed to host the component library. Additionally, it defines the process by which users request access for the contribution and retrieval of library content.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

Neutron and Thermal Embrittlement of RPV Steels: An Overview

Because the reactor pressure vessel (RPV) represents the first structural line of defense against the release of radiation to the public, the design and fabrication of the RPV for any nuclear reactor facility is performed at very high standards in accordance with consensus codes that are based on mechanical and physical properties of the steels used to construct the vessel. Nuclear RPVs may weigh up to 800 tons with wall thicknesses up to approximately 330 mm and are clad on the inside with stainless-steel weld metal and given a final post-weld heat treatment. The RPV is a unique structural component in that it operates under high pressures and temperatures and is exposed to relatively high neutron radiation. Although typical RPV steels and welds have excellent fracture toughness at room temperature and above when put into service, the degrading effects of high-energy neutron irradiation can cause levels of irradiation-induced embrittlement in radiation-sensitive materials of concern for the structural integrity of the RPV. In recent decades, remarkable progress has been made in developing a mechanistic understanding of irradiation embrittlement. This progress includes developing physically based and statistically calibrated models of Charpy V-notch-indexed transition temperature shifts based on results from RPV surveillance programs complemented by significant results from comprehensive research experiments performed in test reactors. In addition, advances in elastic-plastic fracture mechanics allow for a relatively small number of relatively small specimens to characterize the fracture toughness of RPV steels with statistical confidence. This paper presents a review of the primary mechanical properties, test procedures, examples of applicable codes and standards, and specimen types used to characterize RPV steels and welds, the effects of neutron irradiation on those most relevant mechanical properties, and a brief review of the effects of thermal aging on RPV materials. The paper closes with a summary.

Nanstad, Randy K.↗
DOE OSTI2017

Validation of the Computational Fluid Dynamics Method using the Aboveground Configuration of the Dry Cask Simulator

Dry cask applicants submit their spent nuclear fuel dry storage and transportation cask designs to the U.S. Nuclear Regulatory Commission (NRC) for certification under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, 'Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste'. The NRC staff performs its technical review of these designs in accordance with 10 CFR Part 72 and NUREG-1536, 'Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,' Revision 1, issued July 2010. To ensure that the cask and fuel material temperatures of the dry cask storage system remain within the allowable limits or criteria for normal, off-normal, and accident conditions, the NRC staff performs a thermal review as part of the technical review. Recent applications increasingly have used thermalhydraulic analyses using computational fluid dynamics (CFD) codes (e.g., ANSYS FLUENT) to demonstrate the adequacy of the thermal design. The NRC performs validation studies of the FLUENT CFD code to assist it in making regulatory decisions to ensure adequate protection for storage and transportation casks. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. In efforts to further validate these designs, the NRC in conjunction with the Department of Energy and Sandia National Laboratories (SNL) are conducting an experiment to capture the effects of elevated helium pressures and accurately portray the external convection of aboveground and belowground canister dry cask systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Overview of the DOE hydrogen safety, codes and standards program, part 3: Advances in research and development to enhance the scientific basis for hydrogen regulations, codes and standards

Hydrogen fuels are being deployed around the world as an alternative to traditional petrol and battery technologies. As with all fuels, regulations, codes and standards are a necessary component of the safe deployment of hydrogen technologies. There has been a focused effort in the international hydrogen community to develop codes and standards based on strong scientific principles to accommodate the relatively rapid deployment of hydrogen-energy systems. The need for science-based codes and standards has revealed the need to advance our scientific understanding of hydrogen in engineering environments. Here, this brief review describes research and development activities with emphasis on scientific advances that have aided the advancement of hydrogen regulations, codes and standards for hydrogen technologies in four key areas: (1) the physics of high-pressure hydrogen releases (called hydrogen behavior); (2) quantitative risk assessment; (3) hydrogen compatibility of materials; and (4) hydrogen fuel quality.

08 HYDROGEN↗
DOE OSTI2024

SCALE Input and Result Files Supporting SCALE Inventory and Reactivity Analysis as Part of the Hermes 2021 PSAR Review

This dataset contains input and result files of computational simulations with the SCALE code system. The simulations cover radionuclide inventory and reactivity analyses of a fluoride salt-cooled high temperature pebble-bed reactor (PB-FHR), specifically the Hermes low-power PB-FHR demonstration reactor. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

equilibrium core↗
DOE OSTI2015

MO-F-CAMPUS-T-02: An Electronic Whiteboard Platform to Manage Treatment Planning Process

Purpose: In an effort to improve patient safety and streamline the radiotherapy treatment planning (TP) process, a software based whiteboard had been developed and put in use in our facility Methods: The electronic whiteboard developed using SQL database (DB) and PHP/JavaScript based web interface, is published via department intranet and login credentials. The DB stores data for each TP process such as patient information, plan type, simulation/start dates, physician, dosimetrist, QA and the current status in planning process. Users interact with the DB per plan and perform status updates in real time as the planning process progresses. All user interactions with the DB are recorded with timestamps so as to calculate statistical information for TP process management such as contouring times, planning and review times, dosimetry, physics and therapist QA times. External beam and brachytherapy plans are categorized according to complexity (ex: IMRT, 3D, HDR, LDR etc) and treatment types and applicators. Each plan category is assigned specific timelines for each planning process. When a plan approaches or passes the predetermined timeline, users are alerted via color coded graphical cues. When certain process items are not completed in time, pre-determined actions are triggered such as a delay in treatment start date. Results: Our institution has been using the electronic whiteboard for two years. Implementation of pre-determined actions based on the statistical information collected by the whiteboard improved our TP process. For example, the average time for normal tissue contouring decreased from 0.73±1.37 to 0.24±0.33 days. The average time for target volume contouring decreased from 3.2±2.84 to 2.37±2.54 days. This increase in efficiency allows more time for quality assurance processes, improving patient safety. Conclusion: The electronic whiteboard has been an invaluable tool for streamlining our TP processes. It facilitates timely and accurate communication between all parties involved in the TP process increasing patient safety.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2019

A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

2013 Building Technologies Office Program Peer Review Report

The 2013 Building Technologies Office Program Peer Review Report summarizes the results of the 2013 Building Technologies Office (BTO) peer review, which was held in Washington, D.C., on April 2–4, 2013. The review was attended by over 300 participants and included presentations on 59 BTO-funded projects: 29 from BTO’s Emerging Technologies Program, 20 from the Commercial Buildings Integration Program, 6 from the Residential Buildings Integration Program, and 4 from the Building Energy Codes Program. This report summarizes the scores and comments provided by the independent reviewers for each project.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2026

Documentation, assessment, and improvements of fission gas modeling capabilities in BISON

Fission gas behavior critically affects nuclear fuel performance, influencing thermal conductivity, microstructure, swelling, and fuel-cladding interactions. The Simple Integrated Fission Gas Release and Swelling (Sifgrs) model, implemented within the BISON fuel performance code, provides advanced capabilities for modeling these phenomena. This manuscript reviews and assesses Sifgrs’ modular structure and submodels describing intragranular, intergranular, and release mechanisms. The modular implementation enables rapid improvements as our understanding advances and new fuel forms emerge. After an overview of capabilities, we present a comprehensive validation suite, comparing predictions to experimental data from tests under normal and transient conditions, including reproduction of the Vitanza curve. Results demonstrate Sifgrs’ accuracy, robustness, and applicability. Current limitations and ongoing efforts are discussed, including mechanistic, multiscale modeling, uncertainty quantification, and validation within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, supporting the safe and efficient use of nuclear fuels.

BISON↗
DOE OSTI2019

Using cylindrical implosions to investigate hydrodynamic instabilities in convergent geometry

Hydrodynamic instabilities such as the Rayleigh–Taylor (RT) and Richtmyer–Meshkov instabilities disrupt inertial confinement fusion (ICF) implosions through the growth of 3D perturbations. Growth of these 3D imperfections at the interfaces of an ICF capsule during implosion lead to mixing between materials that is detrimental to performance. These instabilities have been studied extensively in planar geometry, but such experiments lack the effects of convergence in spherical implosions. While several studies have been performed in spherical geometry, these often lack a direct means to measure perturbation growth. Experiments in cylindrical geometry include convergence effects while maintaining direct diagnostic access. Although cylinders have less compression than spheres, they do provide an excellent platform to validate modeling for convergent geometries. The problem with previous cylindrical implosion experiments was that the convergence ratios were limited to ~4. With the National Ignition Facility (NIF), larger cylindrical targets can be driven to convergences of 10–15 while maintaining a large enough final diameter to measure perturbation growth. This paper reviews the design process used to both benchmark radiation hydrodynamics codes and enable 1D post-processed simulations to explore design space to separate compression effects from acceleration/deceleration RT instability. Results from 1D simulations suggest that cylindrical implosions on the NIF can produce high-convergence experiments to validate RT instability growth for ICF implosions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Code for paper "Reconstructing an Implosion with Space-Time Smoothing of Velocimetry Traces"

New diagnostics to obtain high-temporal resolution observation of surface velocity provide a rich collection of data from high-explosive (HE) experiments. Using this data to estimate the temporal evolution of an imploding, HE-driven surface is of great interest for validating simulation of the physical system and for design of future experiments. This is the code for a paper in which we propose a method for using a Gaussian process to estimate the surface position over time and provide uncertainty estimates on that reconstruction accounting for measurement error, uncertainty to sparse spatial sampling velocity data, and robustness to outlying observations. In the paper, the code is applied on data from a hemispherical HE implosion experiment. This code is for release with the paper for peer-review, replicating the results and making the figures from the paper.

Grosskopf, Michael↗