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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 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Practical overview of the ICRP Publication 145 phantoms for use with MCNP

Abstract This paper provides a practical overview of the mesh-type reference phantoms (MRCPs) published by the ICRP by providing supplementary guidance on the proper implementation of using the MRCPs in computational codes, such as Monte Carlo N-Particle. Various validation scenarios for photon and neutron exposures were also examined where published values were reproduced using the MRCPs for both external and internal exposures. Generally, the MRCPs performed well in the validation scenarios with the values reproduced being within 10% of the published values.

Hetrick, Lucas (ORCID:0000000178978007)↗
DOE OSTI2019

The Effects of Radiation and Emitted Light Transport on the Positional Response of 11 cm × 42.5 cm × 5.5 cm NaI(Tl) Detectors

Experiments were performed with 30 11 cm x 42.5 x cm 5.5 cm NaI(Tl) detectors to better understand their positional response. Spectra were collected using 0.02 to 0.15 MBq point sources of 241 Am, 137 Cs, 60 Co, and 133 Ba positioned on lines parallel and perpendicular to the long axis of the crystal along both the narrow and wide detector faces as well as at different distances from them. A greater density of positions was sampled at the ends of the detector, and repeated measurements were made to examine potential gain drifts during the experiment. Spectroscopic peak counts, spectroscopic pulse heights, and net counts were analyzed. Empirical equations were fit to the aforementioned data for each specific source energy as a function of source position. In addition, a Monte Carlo radiation transport code was used to simulate the expected positionally variable response based solely upon radiation absorption. The simulated radiation transport efficiency functions were compared to the experimental data. The effects of the geometric radiation efficiency, the attenuation and scattering of emitted light within the scintillation crystal, and combined effects such as non-uniformity of the photomultiplier tube photocathode response and crystal irregularities were then distinguished. Functions describing each effect were derived. The results suggest potential new corrections to data obtained with large scintillation detectors as well as a novel approach to partial positional gamma-ray detection with minimal collimation, given that the energy resolution is within reason for particular photopeaks.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2018

Neutron-neutron angular correlations in spontaneous fission of Cf 252 and Pu 240

Background: Angular anisotropy has been observed between prompt neutrons emitted during the fission process. Such an anisotropy arises because the emitted neutrons are boosted along the direction of the parent fragment. Purpose: To measure the neutron-neutron angular correlations from the spontaneous fission of 252 Cf and 240 Pu oxide samples using a liquid scintillator array capable of pulse-shape discrimination. To compare these correlations to simulations combining the Monte Carlo radiation transport code MCNPX with the fission event generator FREYA. Method: Two different analysis methods were used to study the neutron-neutron correlations with varying energy thresholds. The first is based on setting a light output threshold while the second imposes a time-of-flight cutoff. The second method has the advantage of being truly detector independent. Results: The neutron-neutron correlation modeled by FREYA depends strongly on the sharing of the excitation energy between the two fragments. The measured asymmetry enabled us to adjust the FREYA parameter x in 240 Pu , which controls the energy partition between the fragments and is so far inaccessible in other measurements. The 240 Pu data in this analysis was the first available to quantify the energy partition for this isotope. The agreement between data and simulation is overall very good for 252 Cf ( sf ) and 240 Pu ( sf ). Conclusions: The asymmetry in the measured neutron-neutron angular distributions can be predicted by FREYA. The shape of the correlation function depends on how the excitation energy is partitioned between the two fission fragments. Lastly, experimental data suggest that the lighter fragment is disproportionately excited.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

CHARM-SYCL & IRIS: A Tool Chain for Performance Portability on Extremely Heterogeneous Systems

Performance portability is becoming crucial as high-performance computing systems become increasingly heterogeneous. We have many options for CPUs and accelerators (e.g., GPUs) but also for non-Von Neumann architectures such as field-programmable gate arrays. This paper presents the CHARM-SYCL unified programming environment for multiple accelerator types as a performance-portable programming environment. It uses the IRIS library developed at Oak Ridge National Laboratory as the back end accelerator runtime. IRIS has a high-performance scheduler to distribute tasks across accelerators. This design allows us to run an application from the same source on multiple systems with multiple configurations. We provide three types of portability with CHARM-SYCL: Portable Workflow, Compiler and Runtime Portability, and Application and Performance Portability. We implement a Monte Carlo simulation benchmark code on the CHARM-SYCL execution environment and demonstrate that our programming environment can accommodate extremely heterogeneous systems.

Fujita, Norihisa↗
DOE OSTI2020

Kinetic Simulations of Microstructural Evolution (KSOME)

"Kinetic simulations of microstructural evolution" (KSOME) is an object Kinetic Monte Carlo (OMKC) simulation code. It is developed to simulate microstructural evolution in materials during irradiation. KSOME is developed from ground up with priority given to the flexibility, ease of upgradability, and computation efficiency, in that order. With previously developed simulation codes, various processes (more precisely, the diffusion-reaction processes) that can occur in a system were hardwired into the code. Hence, to perform higher fidelity simulations, a significant coding effort was required every time to incorporate either a more accurate description of existing processes or new processes. With KSOME, the necessity for frequent code upgrades whenever there is a need to perform higher fidelity simulations is either eliminated or reduced significantly. In KSOME, the categories of various diffusion-reaction processes and the data-management system are hardwired, while the execution of the diffusion-reaction processes specific to a system is specified via input files. In the present version of KSOME, the analytical expression required to calculate defect interaction radii are inputted via a text file. However, the ability to parse mathematical expressions from a text file can be extended to input the analytical expression to calculate other defect properties; including the expressions to calculate the change in the migration barriers of mobile defects due to long-range interaction with extended defects

Nandipati, Giridhar [Pacific Northwest National La↗
DOE OSTI2024

SuperNu Version 4.x

We seek to release SuperNu, Version 4.x, as a continuation of development for the open source SuperNu software. The SuperNu, Version 3.x Monte Carlo radiative transfer code for astrophysical transients was released with GPLv3 copyright, asserted by LANL in 2015. For the next release we have features planned for development , including: opacity implementation (including non-local thermodynamic equilibrium effects), generalized source implementation (e.g. for emulating shock heating in Type II supernovae), 3T (electron, ion, radiation) internal energy update, special relativity corrections through O(v^2/c^2), light polarization (e.g. for comparison to spectroplarimetry observations of supernovae and kilonovae), and infrastructure features (checkpoint and restart of simulations, and tools including setup and Slurm batch scripts and simulation post-processing/analysis scripts). These features are intended to improve the fidelity and/or better understand uncertainty in supernova and kilonova light curve calculations.

Wollaeger, Ryan↗
DOE OSTI2025

Microreactor Optimization Using Simulation And Economics (mouse)

Microreactor Optimization Using Simulation and Economics (MOUSE) is a tool that integrates both nuclear microreactor design and reactor economics to provide comprehensive evaluations and optimizations. This tool enables stakeholders to explore the interplay between technical and economic variables, guiding them towards effective and competitive microreactor solutions. For the reactor core simulations, MOUSE leverages the OpenMC Monte Carlo Particle Transport Code to perform detailed core simulations for various microreactor designs. The included OpenMC models are 2D core designs of a Liquid Metal Thermal Microreactor (LMTR), a Gas-Cooled TRISO-Fueled Microreactor (GCMR), and a Heat Pipe Microreactor. Beyond core design, MOUSE includes simplified calculations for: - Calculating the masses of heat exchangers within the system. - Mechanical power of pumps. - Estimating the area occupied by various buildings within the nuclear plant. For the economic analysis, MOUSE provides detailed bottom-up cost estimates, encompassing a wide range of costs including preconstruction costs, direct costs, indirect costs, training costs, financial costs, operation & maintenance (O&M) costs, and fuel costs. These cost estimations are developed using data from the MARVEL project and additional literature sources, enabling the calculation of total capital costs and levelized cost of energy for both first-of-a-kind and nth-of-a-kind microreactors. MOUSE also enables analysis of the cost drivers and competitiveness in the electricity market. MOUSE allows users to modify a wide array of technical and economic parameters to evaluate different scenarios and their impacts. Examples of these parameters include: Fuels, coolants, or reflector materials Enrichment levels Control drum materials and geometry Fuel pin geometry and materials Moderator pin geometry and materials Reactor core and reflector dimensions Packing factor for the TRISO particles Nuclear reactor power and reactor burnup Number of sensors Shielding thickness Reactor vessel and guard vessel dimensions Operational staff requirements Number of emergency shutdowns Levelization period Interest rate Construction duration Since MOUSE is powered by the WATTS toolkit, it supports optimization studies, parametric analyses, and uncertainty calculations/propagation. The optimization techniques enable users to identify optimal design and economic configurations. The parametric analysis tools allow users to explore the sensitivity of various parameters, while uncertainty propagation helps quantify the impact of uncertainties on overall performance and cost. User Interface and Workflow: Currently, MOUSE is a command-line-based tool. Users can input various reactor design or economic parameters, modify the designs, run simulations, and visualize results through comprehensive data visualization and reporting capabilities. The typical workflow involves setting up the reactor model, defining economic parameters, running simulations, and analyzing the results to make informed decisions. By combining advanced design calculations with detailed economic modeling, MOUSE provides a robust framework for optimizing nuclear microreactor technologies, enhancing their competitiveness, and guiding stakeholders towards innovative and cost-effective solutions.

Hanna, Botros [Idaho National Laboratory (INL), Id↗
DOE OSTI2024

Integrating ytopt and libEnsemble to autotune OpenMC

Ytopt is a Python machine-learning-based autotuning software package developed within the ECP PROTEAS-TUNE project. The ytopt software adopts an asynchronous search framework that consists of sampling a small number of input parameter configurations and progressively fitting a surrogate model over the input-output space until exhausting the user-defined maximum number of evaluations or the wall-clock time. libEnsemble is a Python toolkit for coordinating workflows of asynchronous and dynamic ensembles of calculations across massively parallel resources developed within the ECP PETSc/TAO project. libEnsemble helps users take advantage of massively parallel resources to solve design, decision, and inference problems and expands the class of problems that can benefit from increased parallelism. In this paper we present our methodology and framework to integrate ytopt and libEnsemble to take advantage of massively parallel resources to accelerate the autotuning process. Specifically, we focus on using the proposed framework to autotune the ECP ExaSMR application OpenMC, an open source Monte Carlo particle transport code. OpenMC has seven tunable parameters some of which have large ranges such as the number of particles in-flight, which is in the range of 100,000 to 8 million, with its default setting of 1 million. Setting the proper combination of these parameter values to achieve the best performance is extremely time-consuming. Therefore, we apply the proposed framework to autotune the MPI/OpenMP offload version of OpenMC based on a user-defined metric such as the figure of merit (FoM) (particles/s) or energy efficiency energy-delay product (EDP) on Crusher at Oak Ridge Leadership Computing Facility. In conclusion, the experimental results show that we achieve the improvement up to 29.49% in FoM and up to 30.44% in EDP.

Autotuning↗
DOE OSTI2022

Studsvik R2 materials test reactor Ad Hoc depletion strategy for the derivation of the fuel isotopic composition of the MPCMIV benchmark

The Ad Hoc Depletion Strategy elaborated by the NINE company, developed in support of the organization of the MPCMIV (Multi-physics Pellet Cladding Mechanical Interaction Validation) benchmark input and output specifications, is presented. This work aims at illustrating the strategy itself and then showing the results obtained with its application over the Studsvik R2 Testing Reactor, which is analyzed in the benchmark. The objective of the application of the strategy is to compute the fuel elements isotopic compositions at the beginning of some core loadings of interest for the benchmark. To this objective, it is necessary to implement first the simulation model of the three single assembly types and perform the infinite lattice depletions, then, to build the full core model and to perform the simulation of the core cycle. All the models and simulations were carried out with the use of the Monte Carlo particle transport code Serpent 2. Finally, the simulations results are assessed against Studsvik isotopic compositions of the fuel elements discharged from the R2 Reactor at the end of the core loading. Several assumptions were necessary during all the steps of the strategy, to overcome the lack of information regarding the core management. For this reason, the solution found at the end of the current analysis may not be completely optimized and further improvements regarding the model assumptions will be tested in a future work. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Brazilian CBP - Technoeconomic analysis data

This data is related to the paper entitled "Techno-economic analysis of sugarcane bagasse and straw conversion into cellulosic ethanol via consolidated bioprocessing". That features the evaluation of sugarcane bagasse and straw conversion to ethanol at stand-alone facilities generating electricity from residues. The following scenarios were evaluated: Conventional, featuring hydrothermal pretreatment, fungal cellulase, and yeast fermentation (current commercial standard); Mid-term consolidated bioprocessing (CBP), relying on bagasse solubilization without pretreatment or cotreatment; and Mature CBP, incorporating cotreatment but no pretreatment and considering significant technological advance of the CBP. Available here are the spreadsheets used for Material and Energy balance calculation, Capital and Operational costs estimation and Cash flow analysis. Also available are the description and python code used for Monte Carlo analysis of the ethanol and capital investment variations. This data can be used as a source to implement other techno-economic analysis in the biorefinary context.

09 BIOMASS FUELS↗
DOE OSTI2018

Incorporation of Spatial Stochastic Variability into Paracousti-UQ

Waves propagating through natural materials such as ocean water encounter spatial variations in material properties that cannot easily be predicted or known in advance. Deterministic wave simulation algorithms must assume that all properties throughout the model space are precisely known. However, a stochastic wave simulation tool can parameterize the material as a stochastic medium with a certain probability distribution and correlation length. This report documents the addition of spatial stochastic variability into Paracousti-UQ, Sandia Geophysics Department's 3-D full waveform acoustic algorithm within stochastic media. The ability of the code to replicate Monte Carlo solutions in 1-D spatially variable media is also evaluated.

58 GEOSCIENCES↗
DOE OSTI2018

Implementation and Validation of Photon Transport in OpenMC

OpenMC is a Monte Carlo particle transport code focused on reactor physics calculations. It stochastically simulates neutrons moving through 3D models using constructive solid geometry and continuous-energy cross sections. Recently, a photon transport capability was added to OpenMC. In reactor simulations, photon transport is needed, for example, to account for photon heating: although the majority of the energy deposited at a fission site comes from the kinetic energy of the fission fragments, a nontrivial portion can be carried away by photons. Therefore, photons produced in neutron reactions must be simulated in order to accurately model energy deposition. OpenMC can now simulate neutron-induced photon production as well as perform pure photon calculations. The four basic photon interactions with matter — coherent (Rayleigh) scattering, incoherent (Compton) scattering, the photoelectric effect, and pair production — are modeled. Secondary processes that can create new photons (i.e., atomic relaxation, electron-positron annihilation, and bremsstrahlung) are also included. The methods implemented here were validated against MCNP6 for a variety of materials and energies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Activation and residual dose rates in the NSLS-II storage ring

It is well known that components in high energy accelerators may get activated and expose workers during maintenance activities. While the potential for activation is higher in positive ion accelerators due to the prolific production of neutrons, certain regions and components in high energy electron accelerators too can be activated enough to create residual dose rates that require careful attention for ALARA compliance. In this report, the activation and residual dose rates from a few materials are studied from various beam loss scenarios ranging from chronic point losses to intermittent real loss scenarios obtained from the operational parameters of the NSLS-II storage ring. FLUKA Monte Carlo radiation transport code is used to model the geometry, the time profile of the beam losses and score the resulting time evolution of the remnant activity and the residual dose rates.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Modeling of Pyroprocessing Hot Cell for Process Monitoring Evaluation

In this report, we examine studies undertaken to determine the applicability of applying measurements with a neutron counter within a pyroprocessing hot cell. The studies focus on the modeling of the hot cell and neutron counter in Monte Carlo N- Particle Code (MCNP) to determine expected radiation effects and applicability of neutron counters. The neutron counter of focus was the High Dose Neutron Detector (HDND) developed at Los Alamos National Laboratory (LANL). We studied the HDND to determine what would be the detector response under normal background conditions with operating process equipment as well as the background’s potential effect on counting applications involving a U/TRU ingot produced by the pyroprocessing electrorefiner. To perform this task, we were required to develop models of the pyroprocessing process cell and unit operations and generate source terms for various different times in the process. The data for the pyroprocessing materials came from SNL’s Safeguards and Security Process Model (SSPM) as part of MPACT’s advanced integration mission. This report will discuss the development of the pyroprocessing process cell and unit operations models in MCNP as well as various simulations to determine the applicability of the HDND for process monitoring efforts in the hot cell.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Enhanced Visualization Capabilities in OpenMC: Report of Activities under a DOE GAIN Voucher

This work was performed under a Cooperative Research and Development Agreement (CRADA) through the U.S. Department of Energy’s Gateway for Accelerated Innovations in Nuclear (GAIN) program in partnership with Oklo Inc., the industry participant. The broad goal of this project was to improve the visualization capabilities of OpenMC related to model development and simulation results. OpenMC is a Monte Carlo particle transport code that is developed and maintained by a broad community of contributors, of which Argonne National Laboratory plays a leading role.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Compendium of Material Composition Data for Radiation Transport Modeling

In 2011, Pacific Northwest National Laboratory (PNNL) produced a document known as the Materials Compendium, or Compendium of Material Composition Data for Radiation Transport Modeling, PNNL 15870, Rev. 1, that contains material information useable for modeling purposes for properties of 372 materials. This information is used in several modeling programs used by the radiological/nuclear community, though it is primarily tailored for the Monte-Carlo-N-Particle code produced by Los Alamos National Laboratory. This new document Revision 2 includes a complete review and update of all materials data and references, addressing discrepancies and changes in materials data or references that have occurred since the first revision, an additional 40 materials have been added, primarily newer detector materials developed since the last revision, and isotopic specificity.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Cost-optimized Automated Variance Reduction for Highly Angle-dependent Radiation Transport Analyses [Dissertation]

Monte Carlo variance-reduction techniques that directly bias particle direction have historically received limited attention despite being useful for many radiation transport applications such as those in which radiation travels through large regions with a low probability of colliding. One such technique is known as DXTRAN (short for deterministic transport) in the MCNP Monte Carlo radiation transport code. Until now, effectively applying DXTRAN in calculations is largely based on empirical observations of computational performance. Optimal DXTRAN parameters are identified through manual iteration. This work develops new mathematical descriptions of the DXTRAN variance-reduction process and demonstrates a new automated variance-reduction method that applies these mathematical formulations to determine the optimal application of DXTRAN in a given problem.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗