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At least 1,351 records · Page 75

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DOE OSTI2019

Semi-Supervised Information Extraction for Cancer Pathology Reports

Pathology reports are a main source of data for cancer surveillance programs. Manual coding of pathology reports is labor-intensive but necessary for obtaining labeled data to train automated information extraction systems. In this study, we investigated semi-supervised deep learning, improving the performance of a multitask information extraction system for automated annotation of pathology reports. We used a set of over 374,000 pathology reports from the Louisiana Tumor Registry and a novel convolutional attention-based auto-encoder. We performed a set of experiments comparing supervised training augmented with unlabeled data at 1%, 5%, 10%, and 50% of the original data size. We also compared the impact of extending text processing to include unlabeled tokens. We find that semi-supervised training consistently improved individual performance with increased micro-averaged F-scores between 0.012 and 0.064 and increased macro-averaged F-scores of up to 0.158. This demonstrates that semantic information learned via unsupervised learning can be used to improve supervised clinical task performance.

Qiu, John X.↗

DOE OSTI2020

Optimization of a Cyclone Using Multiphase Flow Computational Fluid Dynamics

The U.S. Department of Energy National Energy Technology Laboratory's (NETL) 50 kW th chemical looping reactor (CLR) has an underperforming cyclone, which was designed using empirical correlations. To improve the performance of this cyclone using computational fluid dynamics (CFD)-based modeling simulations, four critical design parameters including the vortex tube radius and length, barrel radius, and the inlet width and height were optimized. Here, NETL's open source multiphase flow with interphase exchange (MFiX) CFD code has been used to model a series of cyclones by systematically varying the geometric design parameters. To perform the optimization process, the surrogate modeling and sensitivity analysis followed by the optimization capability in nodeworks was used. The basic methodology for the process is to employ a statistical design of experiments (DOE) method to generate sampling simulations that fill the design space. Corresponding CFD models are then created, executed, and postprocessed. A response surface is created to characterize the relationship between input parameters and the quantities of interest (QoI). Finally, the CFD-surrogate is used by an optimization method to find the optimal design condition based on the objective and constraints prescribed. The resulting optimal cyclone has a larger diameter and longer vortex tube, a larger diameter barrel, and a taller and narrower solids inlet. The improved design has a predicted pressure drop 11 times lower than the original design while reducing the mass loss by a factor of 2.3.

42 ENGINEERING↗

DOE OSTI2024

Energy Exascale Computational Fluid Dynamics Simulations With the Spectral Element Method

Development and application of the open-source GPU-based fluid-thermal simulation code, NekRS, are described. Time advancement is based on an efficient kth-order accurate timesplit formulation coupled with scalable iterative solvers. Spatial discretization is based on the high-order spectral element method (SEM), which affords the use of fast, low-memory, matrix-free operator evaluation. Further, recent developments include support for nonconforming meshes using overset grids and for GPU-based Lagrangian particle tracking. Results of large-eddy simulations of atmospheric boundary layers for wind-energy applications as well as extensive nuclear energy applications are presented.

42 ENGINEERING↗

DOE OSTI2014

SU-E-T-350: Verification of Gating Performance of a New Elekta Gating Solution: Response Kit and Catalyst System

Purpose: In this work, we have tested the performance of new respiratory gating solutions for Elekta linacs. These solutions include the Response gating and the C-RAD Catalyst surface mapping system.Verification measurements have been performed for a series of clinical cases. We also examined the beam on latency of the system and its impact on delivery efficiency. Methods: To verify the benefits of tighter gating windows, a Quasar Respiratory Motion Platform was used. Its vertical-motion plate acted as a respiration surrogate and was tracked by the Catalyst system to generate gating signals. A MatriXX ion-chamber array was mounted on its longitudinal-moving platform. Clinical plans are delivered to a stationary and moving Matrix array at 100%, 50% and 30% gating windows and gamma scores were calculated comparing moving delivery results to the stationary result. It is important to note that as one moves to tighter gating windows, the delivery efficiency will be impacted by the linac's beam-on latency. Using a specialized software package, we generated beam-on signals of lengths of 1000ms, 600ms, 450ms, 400ms, 350ms and 300ms. As the gating windows get tighter, one can expect to reach a point where the dose rate will fall to nearly zero, indicating that the gating window is close to beam-on latency. A clinically useful gating window needs to be significantly longer than the latency for the linac. Results: As expected, the use of tighter gating windows improved delivery accuracy. However, a lower limit of the gating window, largely defined by linac beam-on latency, exists at around 300ms. Conclusion: The Response gating kit, combined with the C-RAD Catalyst, provides an effective solution for respiratorygated treatment delivery. Careful patient selection, gating window design, even visual/audio coaching may be necessary to ensure both delivery quality and efficiency. This research project is funded by Elekta.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

TH-CD-201-10: Highly Efficient Synchronized High-Speed Scintillation Camera System for Measuring Proton Range, SOBP and Dose Distributions in a 2D-Plane

Purpose: Proton therapy (PT) delivery is complex and extremely dynamic. Therefore, quality assurance testing is vital, but highly time-consuming. We have developed a High-Speed Scintillation-Camera-System (HS-SCS) for simultaneously measuring multiple beam characteristics. Methods: High-speed camera was placed in a light-tight housing and dual-layer neutron shield. HS-SCS is synchronized with a synchrocyclotron to capture individual proton-beam-pulses (PBPs) at ∼504 frames/sec. The PBPs from synchrocyclotron trigger the HS-SCS to open its shutter for programmed exposure-time. Light emissions within 30×30×5cm3 plastic-scintillator (BC-408) were captured by a CCD-camera as individual images revealing dose-deposition in a 2D-plane with a resolution of 0.7mm for range and SOBP measurements and 1.67mm for profiles. The CCD response as well as signal to noise ratio (SNR) was characterized for varying exposure times, gains for different light intensities using a TV-Optoliner system. Software tools were developed to analyze ∼5000 images to extract different beam parameters. Quenching correction-factors were established by comparing scintillation Bragg-Peaks with water scanned ionization-chamber measurements. Quenching corrected Bragg-peaks were integrated to ascertain proton-beam range (PBR), width of Spared-Out-Bragg-Peak (MOD) and distal.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNF system designs (e.g., storage, transfer, transport), SNF characteristics (e.g., fuel types, burnup, cooling time), and loading maps (e.g., uniform loading, zone loading), while a constant decay heat was maintained. The decay heat was calculated using US Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54, Revision 2, and verified using ORIGEN sequence within SCALE code system. The ORIGEN outputs were used as source terms in the dose analysis using 198 different configurations. A computer script was developed to calculate the cooling time necessary to achieve a given decay heat for a given enrichment, assembly average burnup, assembly mass, and in-core history using a rootfinder algorithm. Initially the computer script was developed to provide cooling time and burnup calculations directly to the analysis of dose rates and decay heats, so a comparison between Used Nuclear Fuel-Storage Transportation and Disposal Analysis Resource Data System (UNF-ST&DARDS) results and RG3.54 data was made; results are included in the appendix to this document. However, an iterative approach was used to compute cooling time, and its accuracy did not depend on the results of the RG3.54r2 algorithm, although the algorithm was still used. The results of the evaluation presented herein clearly demonstrate that a given decay heat does not correspond to a unique dose rate for a variety of cask and package designs. There is no clear pattern to develop a correlation between decay heat and the source terms. Depending on burnup, enrichment, cask type, and loading pattern, dose rates varied for the exact same decay heat—in some cases by 400% for a given cask. For cases in which decay heat was held constant through selection of the appropriate cooling time, dose rates would decrease with increasing burnup, and in other cases, dose rates would increase. The large variation in dose rates for a constant decay heat indicates that casks loaded based on decay heat—that is allowing any burnup, cooling time, and enrichment combinations that yield the qualified decay heat limit(s) —cannot ensure that an Independent Spent Fuel Storage Installation or a spent fuel transportation package will meet the regulatory limits set forth by the respective regulations, i.e., 10 CFR 72 or 10 CFR 71.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Alternative Source Verification for the NB5 Beamline

This report details the calculation and comparison of secondary neutron sources for the purpose of performing NB5 velocity selector shielding calculations without relying on the primary HB4 source and the use of a modified MCNP code to run the full model.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

High-energy Gamma Rays from the Milky Way: Three-dimensional Spatial Models for the Cosmic-Ray and Radiation Field Densities in the Interstellar Medium

High-energy γ -rays of interstellar origin are produced by the interaction of cosmic-ray (CR) particles with the diffuse gas and radiation fields in the Galaxy. The main features of this emission are well understood and are reproduced by existing CR propagation models employing 2D galactocentric cylindrically symmetrical geometry. However, the high-quality data from instruments like the Fermi Large Area Telescope reveal significant deviations from the model predictions on few to tens of degrees scales, indicating the need to include the details of the Galactic spiral structure and thus requiring 3D spatial modeling. In this paper, the high-energy interstellar emissions from the Galaxy are calculated using the new release of the GALPROP code employing 3D spatial models for the CR source and interstellar radiation field (ISRF) densities. Three models for the spatial distribution of CR sources are used that are differentiated by their relative proportion of input luminosity attributed to the smooth disk or spiral arms. Two ISRF models are developed based on stellar and dust spatial density distributions taken from the literature that reproduce local near- to far-infrared observations. The interstellar emission models that include arms and bulges for the CR source and ISRF densities provide plausible physical interpretations for features found in the residual maps from high-energy γ -ray data analysis. The 3D models for CR and ISRF densities provide a more realistic basis that can be used for the interpretation of the nonthermal interstellar emissions from the Galaxy.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2016

WEC-Sim Wave Energy Converter Simulator

WEC-Sim (Wave Energy Converter SIMulator) is an open-source wave energy converter (WEC) simulation tool. The code is developed in MATLAB/SIMULINK using the multi-body dynamics solver SimMechanics. WEC-Sim has the ability to model devices that are comprised of rigid bodies, power-take-off systems, and mooring systems. Simulations are performed in the time-domain by solving the governing WEC equations of motion in 6 degrees-of-freedom. The WEC-Sim project is funded by the U.S. Department of Energy's Wind and Water Power Technologies Office and the code development effort is a collaboration between the National Renewable Energy Laboratory (NREL) and Sandia National Laboratories (SNL).

13 HYDRO ENERGY↗

DOE OSTI2018

Mode Conversion Losses in Expansion Units for ITER ECH Transmission Lines

The ITER electron cyclotron heating transmission lines will consist of 63.5 mm diameter corrugated waveguides, each carrying 1 MW of 170 GHz microwaves. These transmission lines must include expansion units to accommodate expansion and contraction along the path from the gyrotron microwave sources to the tokamak. A numerical mode matching code has been developed to calculate power losses due to mode conversion of the operating mode, HE11, to higher order modes as a result of the radial discontinuities in a sliding joint. Two expansion unit designs were evaluated, a simple gap expansion unit and a more complex tapered expansion unit. The gap expansion unit demonstrated loss that oscillated rapidly with expansion length, due to trapped modes within the unit. The tapered expansion unit has been shown to effectively suppress these trapped modes at the expense of increased fabrication complexity. In a gap expansion unit, for a waveguide step size of 2.5 mm, loss can be kept below 0.1% to a maximum expansion length of 17 mm. Expansion units without corrugation on interior walls were also evaluated. Expansion units that lack corrugations are found to increase mode trapping within the units, though not beyond useful application. The mode matching code developed in this paper was also used to estimate mode conversion loss in vacuum pumpouts for the ECH lines; the estimated loss was found to be negligibly small.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Multi-Bunch Stability Analysis of the Advanced Photon Source Upgrade Including the Higher-Harmonic Cavity

Multi-bunch stability simulations were done for the very-low-emittance hybrid seven-bend-achromat (H7BA) lattice proposed for the Advanced Photon Source (APS) upgrade. The simulations, performed using tracking code elegant, were meant to determine whether the long-term wakefields of the higher-order modes (HOMs) of the main 352-MHz cavities will produce an instability. The multi-particle simulations include the important effects of the Higher-Harmonic Cavity (HHC) and the longitudinal impedance of the new vacuum chamber. These realistic simulations show that the HHC provides additional damping in the form of the Landau damping. Still, the HOMs may likely produce a multi-bunch instability which can be cured with more effective HOM damping or a longitudinal feedback system.

Emery, L.↗

DOE OSTI2018

Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Thermodynamic Limits of Redox-Based Thermochemical Processes (REDOTHERM)

Solar thermochemical fuel production is a potential pathway for the production of sustain liquid drop-in fuels, which can help decarbonize the aviation and maritime sectors. In an attempt to analyze the commercial viability of this technology, several studies have been conducted, including system and technoeconomic analysis (TEA) modeling. However, most studies to date simply assume a given redox reactor efficiency, which is significantly higher than demonstrated values to date. While it is widely recognized that utilizing a counter-current flow (CF) configuration could increase the redox reactor efficiency, an over-simplification in the thermodynamic modeling may lead to unphysical results which has been included in multiple publications. The fact that the solar redox reactor is the least developed component in the process chain makes it hard to identify technology gaps and evaluate pathways to deployment at scale using this approach. In this work, a thermodynamic model for a moving oxide system has been developed, in a general form that allows to analyze the system for different redox-active materials, under a wide range of operating conditions, for both parallel and countercurrent flows. The model capabilites are demonstrated, and the model's code will be shared as an open-source on GitHub in the next few months.

chemical looping↗

DOE OSTI2024

Development of The DOME Shield Model For The NRIC Virtual Test Bed

As several advanced reactor concepts are maturing, test beds are needed to accelerate the demonstration and deployment of these advanced nuclear technologies. The National Reactor Innovation Center (NRIC) is building new or enhancing existing US Department of Energy infrastructure to support testing of components and systems. Demonstration of Microreactor Experiments (DOME) will utilize the Experimental Breeder Reactor-II (EBR-II) dome containment structure to host reactor demonstrations. A reactor supplemental shielding is needed so that DOME dose requirements are met. To accelerate the confirmatory analysis required for the reactor demonstration, the NRIC Virtual Test Bed (VTB) is developing a virtual model of the DOME shield that will be made available on the VTB public repository. This will allow developers to leverage advanced modeling and simulation tools to ensure their reactor demonstration concept will meet dose requirement and the limit concrete temperature in the shield during steady state and transient operation conditions. This paper presents the model developed for the DOME shield using open-source tools: MOOSE heat transfer module, Monte Carlo code OpenMC, and Cardinal to calculate the DOME shield temperature distribution during steady state

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Gamma ray Imaging Based Methods to Improve the Accuracy of Uranium Holdup Quantification

Deposits in commercial uranium processing facilities of interest to nuclear safeguards are predominantly low enriched uranium (LEU). The shapes and sizes of holdup deposits can vary significantly, thus making it difficult to model the deposits accurately for calibration purposes. This presents a challenge to currently employed methods such as the Generalized Geometry Holdup (GGH) that rely on quantifying 235U mass by simplifying deposit shapes as a point, a line, or an area. In this work, gamma ray imaging using high energy resolution Germanium Gamma Imagers (GeGI™) is employed to determine the distribution of uranium inside the source containment. Two types of imaging methods are employed; coded-aperture imaging and Compton imaging. Gamma-ray emissions seen in different regions of the image are quantified using an inverse gamma-ray transport solver being developed at ORNL. Using measured data, the inverse method will be used to solve for unknown source parameters such as source matrix thickness, density, and attenuation due to container wall and shielding. The intrinsic efficiency of the detector is determined based on a library of built-in response functions and key dimensions of the detector (e.g., thickness and radius of the detector crystal). Measurements are performed in both imaging modalities using uranium sources of well-known masses and enrichments configured inside mocked-up holdup fixtures. Results from the two imaging modalities are intercompared. The paper presents preliminary results of the imaging measurements and the progress made in the computational methods to quantify uranium mass.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2023

Code-verification techniques for the method-of-moments implementation of the combined-field integral equation

Code verification plays an important role in establishing the credibility of computational simulations by assessing the correctness of the implementation of the underlying numerical methods. In computational electromagnetics, the numerical solution to integral equations incurs multiple interacting sources of numerical error, as well as other challenges, which render traditional code-verification approaches ineffective. In this paper, we provide approaches to separately measure the numerical errors arising from these different error sources for the method-of-moments implementation of the combined-field integral equation. Here, we demonstrate the effectiveness of these approaches for cases with and without coding errors.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

Approach for the isotopic characterization of irradiated start-up sources for disposal at WIPP - 15674

The National Nuclear Security Administration, Global Threat Reduction Initiative (GTRI)/ Off-site Source Recovery Program (OSRP) requires an approved final disposition pathway for sealed irradiated startup sources with uncertain irradiation histories. To create this path a reliable method is required to determine yields of transmutation, fission and activation products contained within the source at the time of disposal. Because of the uncertain histories of these sources, a number of likely histories are modeled with the code ORIGEN in order to establish the bounds of the radioisotope yields. The ORIGEN simulations are confirmed with gamma spectroscopy, specifically examining the activities of several fission and transmutation isotopes. These calculations, coupled with other previously established characterization techniques for transuranic neutron sources, will be used to establish a waste profile that is in compliance with Waste Isolation Pilot Plant (WIPP) waste acceptance criteria. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Cross Section Generation Capability in Griffin

The Griffin code is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory. The code includes a variety of steady-state solvers for fixed-source, k-eigenvalue, adjoint, and subcritical multiplication, as well as transient solvers for point-kinetics, improved quasi-static, and spatial dynamics. The code reads multigroup cross sections in the ISOXML format generated from external deterministic or Monte Carlo cross section generation codes. The implementation of the cross section generation capability in Griffin was initiated last year by plugging in the cross section application programming interface (CSAPI) and reviewing the methodologies for treating particulate fuels. The focus this year was on improving the CSAPI integration and implementing advanced self-shielding methods for applications to advanced reactor problems with TRISO fuels. First, the process for cross section library generation was updated to accurately and rigorously produce isotopic cross section data. Second, the equivalent Dancoff factor cell method performing slowing down calculations on the fly for the resonance treatment was implemented in CSAPI to improve the accuracy of effective multigroup cross sections in the resonance energy range. Third, the iterative local spatial self-shielding method was implemented under the calculation framework of the equivalent Dancoff factor cell method to accurately deal with the double heterogeneity effect of particulate fuel. The updated CSAPI with the advanced self-shielding methods, together with the cross section libraries generated based on the improved process, were tested for pin-cell, unit-cell, and fuel assembly problems with various resonance self-shielding conditions based on very high temperature reactor, high temperature test reactor, and Empire benchmark cores, indicating that the updated CSAPI in Griffin is able to produce multigroup cross sections accurately and efficiently. We also showed that the methodology worked well for pebble bed fuel from HTR-10, but the capability still needs to be fully integrated into CSAPI. In the future, further benchmark tests will be performed for various thermal reactor core problems, including particulate fuel-based pebble bed reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗