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DOE OSTI2019

Status of the NEAMS and ARC neutronic fast reactor tools integration to the NEAMS Workbench

The Workbench initiative was launched in FY-2017 within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Integration Product Line to facilitate the transition from conventional tools to high-fidelity tools. The Workbench provides a common user interface for model creation, real-time validation, execution, output processing, and visualization for integrated codes. The integration of the Argonne Reactor Computation (ARC) suite of codes into the NEAMS Workbench was initiated in FY-2017. The ARC codes contain both legacy codes like DIF3D and REBUS-3 that were developed with over 30 years of experience, and newer NEAMS additions like MC2-3, PERSENT and PROTEUS. The ARC integration into the NEAMS Workbench interface relies on the PyARC module which handles the pre- and post-processing of the native ARC codes input, and the runtime environment. The PyARC module together with the NEAMS Workbench interface are both released under Open Source Software licenses. In FY-2019, the effort focused on integrating the explicit assembly management logic of REBUS, the GAMSOR code for gamma heating transport, and of the NODAL solver of PROTEUS. Additional capabilities were also integrated in response to user requests, such as the assembly power peaking calculation and a 2D result plotting capability. For verification and demonstration purposes, the ARC codes were used through the Workbench for solving the Sodium-cooled Fast Reactor Uncertainty Analysis in Modelling (SFR-UAM) benchmark problems, including the newly proposed unit cells and ASTRID problem. Significant effort in FY-2019 focused on training new users from ANL, INL, NCSU, and Westinghouse. In particular, extensive training material was developed in the form of tutorials, including documentation, sample inputs, and associated presentations. The ARC codes are now actively used through the NEAMS Workbench by nuclear engineers at ANL, INL, NCSU, and Westinghouse, for LFR, MSR, micro-reactor, and SFR core design analyses. Future efforts will focus on adding new and existing modeling capabilities available with the ARC and NEAMS codes, training new users and supporting them to continue building user experience.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2026

Accelerating lattice gauge theory studies with Agentic AI

Lattice gauge theory research, with its computationally intensive simulations and complex multi‑stage workflows, is well positioned to benefit from agentic AI systems. We demonstrate how such tools can support key components of lattice gauge theory research, including novel simulation code development using standard LQCD frameworks, HPC job orchestration, simulation data analysis, and expert‑guided tuning of algorithmic parameters such as Hasenbusch mass preconditioning and multigrid solvers. Our results show that agentic AI can reduce manual effort, improve productivity, and accelerate the research cycle while maintaining essential human oversight.

Ayyar, Venkitesh [Fermilab]↗

DOE OSTI2016

ecode - Electron Transport Algorithm Testing v. 1.0

ecode is a Monte Carlo code used for testing algorithms related to electron transport. The code can read basic physics parameters, such as energy-dependent stopping powers and screening parameters. The code permits simple planar geometries of slabs or cubes. Parallelization consists of domain replication, with work distributed at the start of the calculation and statistical results gathered at the end of the calculation. Some basic routines (such as input parsing, random number generation, and statistics processing) are shared with the Integrated Tiger Series codes. A variety of algorithms for uncertainty propagation are incorporated based on the stochastic collocation and stochastic Galerkin methods. These permit uncertainty only in the total and angular scattering cross sections. The code contains algorithms for simulating stochastic mixtures of two materials. The physics is approximate, ranging from mono-energetic and isotropic scattering to screened Rutherford angular scattering and Rutherford energy-loss scattering (simple electron transport models). No production of secondary particles is implemented, and no photon physics is implemented.

Franke, Brian C.↗

DOE OSTI2014

TRIGA MARK-II source term

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

A scintillator-based approach to monitor secondary neutron production during proton therapy

Here, the primary objective of this work is to measure the secondary neutron field produced by an uncollimated proton pencil beam impinging on different tissue-equivalent phantom materials using organic scintillation detectors. Additionally, the Monte Carlo code mcnpx-PoliMi was used to simulate the detector response for comparison to the measured data. Comparison of the measured and simulated data will validate this approach for monitoring secondary neutron dose during proton therapy. Proton beams of 155- and 200-MeV were used to irradiate a variety of phantom materials and secondary particles were detected using organic liquid scintillators. These detectors are sensitive to fast neutrons and gamma rays: pulse shape discrimination was used to classify each detected pulse as either a neutron or a gamma ray. The mcnpx-PoliMi code was used to simulate the secondary neutron field produced during proton irradiation of the same tissue-equivalent phantom materials. As a result, an experiment was performed at the Loma Linda University Medical Center proton therapy research beam line and corresponding models were created using the mcnpx-PoliMi code. The authors’ analysis showed agreement between the simulations and the measurements. The simulated detector response can be used to validate the simulations of neutron and gamma doses on a particular beam line with or without a phantom. In conclusion, the authors have demonstrated a method of monitoring the neutron component of the secondary radiation field produced by therapeutic protons. The method relies on direct detection of secondary neutrons and gamma rays using organic scintillation detectors. These detectors are sensitive over the full range of biologically relevant neutron energies above 0.5 MeV and allow effective discrimination between neutron and photon dose. Because the detector system is portable, the described system could be used in the future to evaluate secondary neutron and gamma doses on various clinical beam lines for commissioning and prospective data collection in pediatric patients treated with proton therapy.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

Modeling the Ex-core Detector Response into a Production Option in MPACT

In the past few years, the Consortium for Advanced Simulation of Light Water Reactors (CASL) has overseen a successful activity to extend the radiation transport capability from the reactor in-core geometry to the ex-core geometry by coupling the deterministic neutronics code MPACT, with the Monte Carlo code Shift. However, transient problems such as RIAs require an efficient prediction of the ex-core detector response to determine the timeframe for a reactor trip. Therefore, in addition to the high-fidelity MPACT-Shift coupled calculation, separate efforts have been made to explore simplified methods for ex-core detector response with MPACT standalone calculation. In a previous milestone L3_RTM.MCH.P17.06, several simplified methods for ex-core detector response were developed in MPACT. In particular, the double kernel method is based on the default MPACT core geometry, so extensive modeling or transport calculation with full reflector and ex-core details are not needed in this method. Since the 1-D diffusion and neutron streaming kernels are used for extended ex-core calculations, the extra computational efforts are very small. The work implemented in the current milestone provides a fully integrated option for MPACT to approximately calculate the ex-core detector response using the double kernel method. To overcome the difficulty of obtaining the consistent diffusion coefficients (these were obtained through separate CMFD calculation and edits), the 1-D diffusion solver is replaced by a 1-D Sn transport solver to compute the neutron flux at the vessel outer surface. The new VERA inputs for ex-core detector are processed in MPACT. The detector signals are linked to the SCRAM logic as an option to trip the core during transient calculation. Two sets of 3-D quarter core problems are run with MPACT standalone and VeraShift to verify the radial and axial effects on detector response by varying the moderator density and control rod position. For the cases with moderator density changes, the detector responses from MPACT are within 2 sigma error of VeraShift. For the cases at different control rod positions, the axial offsets computed from MPACT are within 2% error of VeraShift results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Magnetar-powered Supernovae in Two Dimensions. II. Broad-line Supernovae Ic

Nascent neutron stars (NSs) with millisecond periods and magnetic fields in excess of 10{sup 16} Gauss can drive highly energetic and asymmetric explosions known as magnetar-powered supernovae. These exotic explosions are one theoretical interpretation for supernovae Ic-BL, which are sometimes associated with long gamma-ray bursts. Twisted magnetic field lines extract the rotational energy of the NS and release it as a disk wind or a jet with energies greater than 10{sup 52} erg over ∼20 s. What fraction of the energy of the central engine go into the wind and the jet remain unclear. We have performed two-dimensional hydrodynamical simulations of magnetar-powered supernovae (SNe) driven by disk winds and jets with the CASTRO code to investigate the effect of the central engine on nucleosynthetic yields, mixing, and light curves. We find that these explosions synthesize less than 0.05 M{sub ⊙} of {sup 56}Ni and that this mass is not very sensitive to central engine type. The morphology of the explosion can provide a powerful diagnostic of the properties of the central engine. In the absence of a circumstellar medium, these events are not very luminous, with peak bolometric magnitudes of M{sub b}∼−16.5 due to low {sup 56}Ni production.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2026

Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT

The U.S. nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. Here, this work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles' silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

The challenges of beam polarization and keV-scale centre-of-mass energy calibration at the FCC-ee

The capability to determine the FCC-ee centre-of-mass energies (ECM) at the ppm level using resonant depolarization of the beams is essential for the Z line shape measurements, the W mass and the possible observation of the Higgs boson s-channel production. A first analysis (Blondel A et al Polarization and centre-of-mass energy calibration at FCC-ee. demonstrated the feasibility of this programme, conditional to careful preparation and a number of further developments. The existing simulation codes must be unified; the analysis and design of the instrumentation must be developed; and a detailed planning must be developed for the simultaneous and coordinated operation of the accelerator, of the continuous polarization and depolarization measurements, and of the beam monitoring devices, ensuring a precise extrapolation from beam energies to centre-of-mass energy and energy spread.

43 PARTICLE ACCELERATORS↗

NASA NTRS2007

Mission Simulation Toolkit

The Mission Simulation Toolkit (MST) is a flexible software system for autonomy research. It was developed as part of the Mission Simulation Facility (MSF) project that was started in 2001 to facilitate the development of autonomous planetary robotic missions. Autonomy is a key enabling factor for robotic exploration. There has been a large gap between autonomy software (at the research level), and software that is ready for insertion into near-term space missions. The MST bridges this gap by providing a simulation framework and a suite of tools for supporting research and maturation of autonomy. MST uses a distributed framework based on the High Level Architecture (HLA) standard. A key feature of the MST framework is the ability to plug in new models to replace existing ones with the same services. This enables significant simulation flexibility, particularly the mixing and control of fidelity level. In addition, the MST provides automatic code generation from robot interfaces defined with the Unified Modeling Language (UML), methods for maintaining synchronization across distributed simulation systems, XML-based robot description, and an environment server. Finally, the MSF supports a number of third-party products including dynamic models and terrain databases. Although the communication objects and some of the simulation components that are provided with this toolkit are specifically designed for terrestrial surface rovers, the MST can be applied to any other domain, such as aerial, aquatic, or space.

Pisaich, Gregory↗

DOE OSTI2025

A Refined BISON-Griffin Workflow For Designing Targeted TRISO Transient Experiments In The Transient Reactor Test Facility

U.S. industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In our previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs,(2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. Predicted tangential stresses in particle silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. Use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Summary report on UO 2 thermal conductivity model refinement and assessment studies

Uranium dioxide (UO 2 ) is the most commonly used fuel in light water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, therefore, governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models were replaced with models that incorporate explicit thermal conductivity degradation mechanisms during fuel burn-up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham type interatomic potential and a potential that combines the many-body embedded atom method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin scattering mechanism due to spins on the magnetic uranium ions have been introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel performance codes. The model is validated by comparison to low-temperature experimental measurements on single crystal hyper-stoichiometric UO 2+x samples and high-temperature literature data. Ongoing works include investigation of the effect of phase separation to UO 2 +U 4 O 9 on the low temperature thermal conductivity of UO 2+x , and modeling of thermal conductivity using the Green-Kubo method. Ultimately, this work will enable more accurate fuel performance simulations as well as extension to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Coupling the System Analysis Module with SAS4A/SASSYS-1

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors, with an emphasis on sodium fast reactors. SAS4A/SASSYS-1 has been under development and in active use for nearly forty-five years, and is currently maintained by the U.S. Department of Energy under the Office of Advanced Reactor Technology. Although SAS4A/SASSYS-1 contains a very capable primary and intermediate system modeling component, PRIMAR-4, it also has some shortcomings: outdated data management and code structure makes extension of the PRIMAR-4 module somewhat difficult. The user input format for PRIMAR-4 also limits the number of volumes and segments that can be used to describe a given system. The System Analysis Module (SAM) is a fairly new code development effort being carried out under the U.S. DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM is being developed with advanced physical models, numerical methods, and software engineering practices; however, it is currently somewhat limited in the system components and phenomena that can be represented. For example, component models for electromagnetic pumps and multi-layer stratified volumes have not yet been developed. Nor is there support for a balance of plant model. Similarly, system-level phenomena such as control-rod driveline expansion and vessel elongation are not represented. This report documents fiscal year 2016 work that was carried out to couple the transient safety analysis capabilities of SAS4A/SASSYS-1 with the system modeling capabilities of SAM under the joint support of the ART and NEAMS programs. The coupling effort was successful and is demonstrated by evaluating an unprotected loss of flow transient for the Advanced Burner Test Reactor (ABTR) design. There are differences between the stand-alone SAS4A/SASSYS-1 simulations and the coupled SAS/SAM simulations, but these are mainly attributed to the limited maturity of the SAM development effort. The severe accident modeling capabilities in SAS4A/SASSYS-1 (sodium boiling, fuel melting and relocation) will continue to play a vital role for a long time. Therefore, the SAS4A/SASSYS-1 modernization effort should remain a high priority task under the ART program to ensure continued participation in domestic and international SFR safety collaborations and design optimizations. On the other hand, SAM provides an advanced system analysis tool, with improved numerical solution schemes, data management, code flexibility, and accuracy. SAM is still in early stages of development and will require continued support from NEAMS to fulfill its potential and to mature into a production tool for advanced reactor safety analysis. The effort to couple SAS4A/SASSYS-1 and SAM is the first step on the integration of these modeling capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Flow Redirection and Induction in Steady State (FLORIS) Wind Plant Power Production Data Sets

This dataset contains turbine- and plant-level power outputs for 252,500 cases of diverse wind plant layouts operating under a wide range of yawing and atmospheric conditions. The power outputs were computed using the Gaussian wake model in NREL's FLOw Redirection and Induction in Steady State (FLORIS) model, version 2.3.0. The 252,500 cases include 500 unique wind plants generated randomly by a specialized Plant Layout Generator (PLayGen) that samples randomized realizations of wind plant layouts from one of four canonical configurations: (i) cluster, (ii) single string, (iii) multiple string, (iv) parallel string. Other wind plant layout parameters were also randomly sampled, including the number of turbines (25-200) and the mean turbine spacing (3D-10D, where D denotes the turbine rotor diameter). For each layout, 500 different sets of atmospheric conditions were randomly sampled. These include wind speed in 0-25 m/s, wind direction in 0 deg.-360 deg., and turbulence intensity chosen from low (6%), medium (8%), and high (10%). For each atmospheric inflow scenario, the individual turbine yaw angles were randomly sampled from a one-sided truncated Gaussian on the interval 0 deg.-30 deg. oriented relative to wind inflow direction. This random data is supplemented with a collection of yaw-optimized samples where FLORIS was used to determine turbine yaw angles that maximize power production for the entire plant. To generate this data, a subset of cases were selected (50 atmospheric conditions from 50 layouts each for a total of additional 2,500 cases) for which FLORIS was re-run with wake steering control optimization. The IEA onshore reference turbine, which has a 130 m rotor diameter, a 110 m hub height, and a rated power capacity of 3.4 MW was used as the turbine for all simulations. The simulations were performed using NREL's Eagle high performance computing system in February 2021 as part of the Spatial Analysis for Wind Technology Development project funded by the U.S. Department of Energy Wind Energy Technologies Office. The data was collected, reformatted, and preprocessed for this OEDI submission in May 2023 under the Foundational AI for Wind Energy project funded by the U.S. Department of Energy Wind Energy Technologies Office. This dataset is intended to serve as a benchmark against which new artificial intelligence (AI) or machine learning (ML) tools may be tested. Baseline AI/ML methods for analyzing this dataset have been implemented, and a link to their repository containing those models has been provided. The .h5 data file structure can be found in the GitHub repository under explore_wind_plant_data_h5.ipynb.

AI↗

DOE OSTI2023

Examining the effects of soil entrainment during nuclear cloud rise on fallout predictions using a multiscale atmospheric modeling framework

Current operational models for nuclear cloud rise over land were developed and validated using observations from shallow-buried or surface detonations, where lofted soil quickly mixed with fission products from the detonation. These models poorly predict fallout from elevated detonations near the fallout-free height of burst (FFHOB), where interactions with the ground are limited and the mixing of fission products and lofted soil is incomplete. Fallout-free is a misnomer at this HOB, as fallout was observed in these cases, but was below the levels of concern, especially off-grounds of the nuclear test site. To correctly characterize and model fallout from detonations near the FFHOB, models must be developed which can capture the stratified nature of the particle and activity-size distributions within the cloud. Previously, it was shown that the Weather Research and Forecasting (WRF) model can accurately simulate nuclear cloud rise for airbursts with little to no ground interactions (Arthur et al., 2021). That work is expanded here by (1) using a radiation-hydrodynamics code to improve the fireball initialization in WRF, (2) further developing an aerosol package from WRF-Chem to simulate lofted soil, and (3) combining the WRF cloud rise simulations with the operational models used at the National Atmospheric Release Advisory Center (NARAC) for fallout modeling. Using this combination of codes, the Upshot-Knothole Grable detonation, which was just below the FFHOB, is simulated from seconds after detonation through cloud rise and fallout, and results are compared to historical test data. Here, the results show improved prediction of dose rate and highlight the need to correctly characterize the entrainment of material into the cloud and the subsequent mixing of fission products with entrained material.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2017

Adverse Events Involving Radiation Oncology Medical Devices: Comprehensive Analysis of US Food and Drug Administration Data, 1991 to 2015

Purpose: Radiation oncology relies on rapidly evolving technology and highly complex processes. The US Food and Drug Administration collects reports of adverse events related to medical devices. We sought to characterize all events involving radiation oncology devices (RODs) from the US Food and Drug Administration's postmarket surveillance Manufacturer and User Facility Device Experience (MAUDE) database, comparing these with non–radiation oncology devices. Methods and Materials: MAUDE data on RODs from 1991 to 2015 were sorted into 4 product categories (external beam, brachytherapy, planning systems, and simulation systems) and 5 device problem categories (software, mechanical, electrical, user error, and dose delivery impact). Outcomes included whether the device was evaluated by the manufacturer, adverse event type, remedial action, problem code, device age, and time since 510(k) approval. Descriptive statistics were performed with linear regression of time-series data. Results for RODs were compared with those for other devices by the Pearson χ{sup 2} test for categorical data and 2-sample Kolmogorov-Smirnov test for distributions. Results: There were 4234 ROD and 4,985,698 other device adverse event reports. Adverse event reports increased over time, and events involving RODs peaked in 2011. Most ROD reports involved external beam therapy (50.8%), followed by brachytherapy (24.9%) and treatment planning systems (21.6%). The top problem types were software (30.4%), mechanical (20.9%), and user error (20.4%). RODs differed significantly from other devices in each outcome (P<.001). RODs were more likely to be evaluated by the manufacturer after an event (46.9% vs 33.0%) but less likely to be recalled (10.5% vs 37.9%) (P<.001). Device age and time since 510(k) approval were shorter among RODs (P<.001). Conclusions: Compared with other devices, RODs may experience adverse events sooner after manufacture and market approval. Close postmarket surveillance, improved software design, and manufacturer–user training may help mitigate these events.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

HARD X-RAY ASYMMETRY LIMITS IN SOLAR FLARE CONJUGATE FOOTPOINTS

The transport of energetic electrons in a solar flare is modeled using a time-dependent one-dimensional Fokker–Planck code that incorporates asymmetric magnetic convergence. We derive the temporal and spectral evolution of the resulting hard X-ray (HXR) emission in the conjugate chromospheric footpoints, assuming thick target photon production, and characterize the time evolution of the numerically simulated footpoint asymmetry and its relationship to the photospheric magnetic configuration. The thick target HXR asymmetry in the conjugate footpoints is found to increase with magnetic field ratio as expected. However, we find that the footpoint HXR asymmetry saturates for conjugate footpoint magnetic field ratios ≥4. This result is borne out in a direct comparison with observations of 44 double-footpoint flares. The presence of such a limit has not been reported before, and may serve as both a theoretical and observational benchmark for testing a range of particle transport and flare morphology constraints, particularly as a means to differentiate between isotropic and anisotropic particle injection.

79 ASTRONOMY AND ASTROPHYSICS↗

NASA NTRS1999

Effect of a Nonplanar Melt-Solid Interface on Lateral Compositional Distribution during Unidirectional Solidification of a Binary Alloy with a Constant Growth Velocity V: Theory - Part 1

Infrared detector materials, such as Hg(1-x)Cd(x)Te, Hg(1-x)Zn(x)Te have energy gaps almost linearly proportional to its composition. Due to the wide separation of liquidus and solidus curves of their phase diagram, there are compositional segregations in both of axial and radial directions of these crystals grown in the Bridgman system unidirectionally with constant growth rate. It is important to understand the mechanisms which affect lateral segregation such that large uniform radial composition crystal is possible. Following Coriell, etc's treatment, we have developed a theory to study the effect of a curved melt-solid interface shape on the lateral composition distribution. The system is considered to be cylindrical system with azimuthal symmetric with a curved melt-solid interface shape which can be expressed as a linear combination of a series of Bessell's functions. The results show that melt-solid interface shape has a dominate effect on lateral composition distribution of these systems. For small values of b, the solute concentration at the melt-solid interface scales linearly with interface shape with a proportional constant of the product of b and (1 - k), where b = VR/D, with V as growth velocity, R as sample radius, D as diffusion constant and k as distribution constant. A detailed theory will be presented. A computer code has been developed and simulations have been performed and compared with experimental results. These will be published in another paper.

Wang, Jai-Ching↗