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At least 1,225 records · Page 68
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

SU-E-T-559: Monte Carlo Simulation of Cobalt-60 Teletherapy Unit Modeling In-Field and Out-Of-Field Doses for Applications in Computational Radiation Dosimetry

Purpose: Mean organ doses from structures located in field and outside of field boundaries during radiotherapy treatment must be considered when looking at secondary effects. Treatment planning in patients with 40 years of follow-up does not include 3-D treatment planning images and did not estimate dose to structures out of the direct field. Therefore, it is of interest to correlate actual clinical events with doses received. Methods: Accurate models of radiotherapy machines combined with whole body computational phantoms using Monte Carlo methods allow for dose reconstructions intended for studies on late radiation effects. The Theratron-780 radiotherapy unit and anatomically realistic hybrid computational phantoms are modeled in the Monte Carlo radiation transport code MCNPX. The major components of the machine including the source capsule, lead in the unit-head, collimators (fixed/adjustable), and trimmer bars are simulated. The MCNPX transport code is used to compare calculated values in a water phantom with published data from BJR suppl. 25 for in-field doses and experimental data from AAPM Task Group No. 36 for out-of-field doses. Next, the validated cobalt-60 teletherapy model is combined with the UF/NCI Family of Reference Hybrid Computational Phantoms as a methodology for estimating organ doses. Results: The model of Theratron-780 has shown to be agree with percentage depth dose data within approximately 1% and for out of field doses the machine is shown to agree within 8.8%. Organ doses are reported for reference hybrid phantoms. Conclusion: Combining the UF/NCI Family of Reference Hybrid Computational Phantoms along with a validated model of the Theratron-780 allows for organ dose estimates of both in-field and out-of-field organs. By changing field size, position, and adding patient-specific blocking more complicated treatment set-ups can be recreated for patients treated historically, particularly those who lack both 2D/3D image sets.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Numerical modeling of the SNS H{sup −} ion source

Ion source rf antennas that produce H- ions can fail when plasma heating causes ablation of the insulating coating due to small structural defects such as cracks. Reducing antenna failures that reduce the operating capabilities of the Spallation Neutron Source (SNS) accelerator is one of the top priorities of the SNS H- Source Program at ORNL. Numerical modeling of ion sources can provide techniques for optimizing design in order to reduce antenna failures. There are a number of difficulties in developing accurate models of rf inductive plasmas. First, a large range of spatial and temporal scales must be resolved in order to accurately capture the physics of plasma motion, including the Debye length, rf frequencies on the order of tens of MHz, simulation time scales of many hundreds of rf periods, large device sizes on tens of cm, and ion motions that are thousands of times slower than electrons. This results in large simulation domains with many computational cells for solving plasma and electromagnetic equations, short time steps, and long-duration simulations. In order to reduce the computational requirements, one can develop implicit models for both fields and particle motions (e.g. divergence-preserving ADI methods), various electrostatic models, or magnetohydrodynamic models. We have performed simulations using all three of these methods and have found that fluid models have the greatest potential for giving accurate solutions while still being fast enough to perform long timescale simulations in a reasonable amount of time. We have implemented a number of fluid models with electromagnetics using the simulation tool USim and applied them to modeling the SNS H- ion source. We found that a reduced, single-fluid MHD model with an imposed magnetic field due to the rf antenna current and the confining multi-cusp field generated increased bulk plasma velocities of > 200 m/s in the region of the antenna where ablation is often observed in the SNS source. We report here on comparisons of simulated plasma parameters and code performance using more accurate physical models, such as two-temperature extended MHD models, for both a related benchmark system describing a inductively coupled plasma reactor, and for the SNS ion source. We also present results from scaling studies for mesh generation and solvers in the USim simulation code.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Physics basis for the Wisconsin HTS Axisymmetric Mirror (WHAM)

The Wisconsin high-temperature superconductor axisymmetric mirror experiment (WHAM) will be a high-field platform for prototyping technologies, validating interchange stabilization techniques and benchmarking numerical code performance, enabling the next step up to reactor parameters. A detailed overview of the experimental apparatus and its various subsystems is presented. WHAM will use electron cyclotron heating to ionize and build a dense target plasma for neutral beam injection of fast ions, stabilized by edge-biased sheared flow. At 25 keV injection energies, charge exchange dominates over impact ionization and limits the effectiveness of neutral beam injection fuelling. This paper outlines an iterative technique for self-consistently predicting the neutral beam driven anisotropic ion distribution and its role in the finite beta equilibrium. Beginning with recent work by Egedal et al. ( Nucl. Fusion , vol. 62, no. 12, 2022, p. 126053) on the WHAM geometry, we detail how the FIDASIM code is used to model the charge exchange sources and sinks in the distribution function, and both are combined with an anisotropic magnetohydrodynamic equilibrium solver method to self-consistently reach an equilibrium. We compare this with recent results using the CQL3D code adapted for the mirror geometry, which includes the high-harmonic fast wave heating of fast ions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Snippet Open-Source Specification. Use and application

Snippet is a collaborative software code auditing tool developed at Sandia National Laboratories (Sandia) in conjunction with several sponsoring government agencies to enable software analysts, individually or in groups, to non-destructively audit large codebases, share files and comments from those audits, and develop Snippet plugins for use with other text editing software applications. This specification defines the requirements and the reasoning behind the requirements for tools developed to be compatible with the technology and purpose of Snippet. A condensation of the requirements is given in the companion document, "Snippet Open-Source Specification Synopsis" also generated by Sandia National Laboratories. Snippet's approach to browsing through codebases is similar to a web browser navigating the internet.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Pedestal particle balance studies in JET-ILW H-mode plasmas

JET-ILW type I ELMy H-modes at 2.5 MA/2.8 T with constant NBI heating (23 MW) and gas fuelling rate were performed, utilising edge localised mode (ELM) pacing by vertical kicks and plasma shaping (triangularity, δ) as tools to disentangle the effects of ELMs, inter-ELM transport and edge stability on the pedestal particle balance. In agreement with previous studies, the pedestal confinement improves with increasing δ, mostly due to a significant increase in pedestal density while the ELM frequency ($f_{\mathrm{ELM}}$) is decreased. Improved pedestal confinement with increasing δ was observed even when the pedestal MHD stability was degraded artificially by vertical kicks, implying that increased triangularity may favourably affect the inter-ELM pedestal recovery. The workflow developed to quantify the pedestal particle balance uses high time-resolution profile reflectometry to characterise the inter-ELM evolution of the plasma particle content ($\mathrm{d} N/\mathrm{d}t$), the NEO drift-kinetic solver to evaluate the neoclassical fluxes and interpretative EDGE2D-EIRENE simulations to estimate the edge particle source. The edge particle source is then constrained by deuterium Balmer-α line intensity measurements in the main chamber, which are, however, strongly affected by reflections from the metal walls. The reflections are accounted for by the CHERAB code taking the divertor emission (the brightest light source in the torus) distribution from imaging spectroscopy measurements as input. Our analysis shows that in the second half of the ELM cycle, the volume-integrated particle source is larger than $\mathrm{d} N/\mathrm{d}t$, indicating that transport plays a key role in the inter-ELM pedestal recovery.

fuelling↗
DOE OSTI2015

Computational Assessment of Naturally Occurring Neutron and Photon Background Radiation Produced by Extraterrestrial Sources

In this study, a computational assessment of the variation in terrestrial neutron and photon background from extraterrestrial sources is presented. The motivation of this assessment is to evaluate the practicality of developing a tool or database to estimate background in real time (or near–real time) during an experimental measurement or to even predict the background for future measurements. The extraterrestrial source focused on during this assessment is naturally occurring galactic cosmic rays (GCRs). The MCNP6 transport code was used to perform the computational assessment. However, the GCR source available in MCNP6 was not used. Rather, models developed and maintained by NASA were used to generate the GCR sources. The largest variation in both neutron and photon background spectra was found to be caused by changes in elevation on Earth's surface, which can be as large as an order of magnitude. All other perturbations produced background variations on the order of a factor of 3 or less. The most interesting finding was that ~80% and 50% of terrestrial background neutrons and photons, respectively, are generated by interactions in Earth's surface and other naturally occurring and man-made objects near a detector of particles from extraterrestrial sources and their progeny created in Earth's atmosphere. In conclusion, this assessment shows that it will be difficult to estimate the terrestrial background from extraterrestrial sources without a good understanding of a detector's surroundings. Therefore, estimating or predicting background during a measurement environment like a mobile random search will be difficult.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

CINDER90 and CINDER2008 Comparison for Second Target Station Analysis

A preliminary target activation analysis for the Second Target Station (STS) at the Spallation Neutron Source (SNS) located at Oak Ridge National Laboratory (ORNL) was preformed using Monte Carlo N-Particle X (MCNPX), CINDER90, and the corresponding multi-group cross section libraries distributed with CINDER90. This preliminary analysis has been used to guide some design decisions of the STS target design. Using the MCNPX output files from this preliminary analysis, a comparison of the activation analysis using CINDER90 and CINDER2008 was performed. While the CINDER90 code has been validated for use in spallation source activation and transmutation calculations, and compares well with measurements, the transition to the modernized version of the code (CINDER2008) and the updated cross section libraries is desirable. The goal of this comparison is to evaluate the differences and implications for the STS activation analysis. Overall, the total activity differs between CINDER2008 and CINDER90 by no more than 20% for the materials evaluated in this report. The largest discrepancies were observed in the decay gamma intensity densities with CINDER90 calculating a factor of 2 higher than CINDER2008.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Code verification for practically singular equations

We report the method-of-moments implementation of the electric-field integral equation (EFIE) yields many code-verification challenges due to the various sources of numerical error and their possible interactions. Matters are further complicated by singular integrals, which arise from the presence of a Green's function. To address these singular integrals, an approach is presented in wherein both the solution and Green's function are manufactured. Because the arising equations are poorly conditioned, they are reformulated as a set of constraints for an optimization problem that selects the solution closest to the manufactured solution. In this paper, we demonstrate how, for such practically singular systems of equations, computing the truncation error by inserting the exact solution into the discretized equations cannot detect certain orders of coding errors. On the other hand, the discretization error from the optimal solution is a more sensitive metric that can detect orders less than those of the expected convergence rate.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

(U) An Analytic Examination of Piezoelectric Ejecta Mass Measurements

Ongoing efforts to validate a Richtmyer-Meshkov instability (RMI) based ejecta source model [1, 2, 3] in LANL ASC codes use ejecta areal masses derived from piezoelectric sensor data [4, 5, 6]. However, the standard technique for inferring masses from sensor voltages implicitly assumes instantaneous ejecta creation [7], which is not a feature of the RMI source model. To investigate the impact of this discrepancy, we define separate “areal mass functions” (AMFs) at the source and sensor in terms of typically unknown distribution functions for the ejecta particles, and derive an analytic relationship between them. Then, for the case of single-shock ejection into vacuum, we use the AMFs to compare the analytic (or “true”) accumulated mass at the sensor with the value that would be inferred from piezoelectric voltage measurements. We confirm the inferred mass is correct when creation is instantaneous, and furthermore prove that when creation is not instantaneous, the inferred values will always overestimate the true mass. Finally, we derive an upper bound for the error imposed on a perfect system by the assumption of instantaneous ejecta creation. When applied to shots in the published literature, this bound is frequently less than several percent. Errors exceeding 15% may require velocities or timescales at odds with experimental observations.

47 OTHER INSTRUMENTATION↗
DOE OSTI2016

Environmental Characterization Scans Using Sodium-Iodide Gamma Spectroscopy to Determine Ra-226 Concentration in Surface Soil - 16326

The recent natural gas drilling boom has brought heightened attention to the issue of TENORM waste management. Radioactive residues from the extraction, treatment, and purification of materials containing uranium and thorium, or their progeny, pose a public health hazard primarily in the form of ground water contamination. Hydraulic fracturing can liberate formation waters from shale; the high salinity of which can cause radium to move with the water to the surface. The disposal and handling of this water has been a subject of public controversy. For soil remediation projects, radium-226 is generally one of the radionuclides of concern due to its alpha emissions. However performing a survey scan for Ra-226 is difficult because its photon emissions are low energy and have minimal probability of emission per decay (186.2 keV, 3.3% abundance). Additionally resolving its emission from U-235 (185.7 keV, 57% abundance) with field instrumentation is not feasible. Therefore detection of Ra-226 relies on the presence and photon emissions of its progeny. This paper outlines methodology for calculating the minimum detectable concentration of Ra-226 in characterization scans using a sodium-iodide detector array coupled with a multi-channel analyzer. Using progeny ingrowth as a detection proxy requires knowing decay time in order to determine progeny emission rates. Knowledge of site history, process, and prior survey results can be key in determining an accurate decay time. However in the absence of such information in-situ measurements can be used to estimate decay time by comparing observed progeny emission ratios against established values. These values are determined using Oak Ridge National Laboratory's ORIGEN-ARP program to generate photon energy spectra for Ra-226 and its progeny at specified time intervals. Once decay time is established, the detector's minimum detectable concentration can be calculated by synthesizing the expected emission spectra and Los Alamos National Laboratory's Monte Carlo N-Particle transport code to simulate detector response. Traditionally, one of the steps in calculating a system's minimum detectable concentration is to place an appropriate button source under the detector, record the response, and then move the source by a specified increment. This process is repeated until the detector's spatial efficiency has been mapped. Due to the impracticality of performing this with Ra-226 button sources of varying decay times, Monte Carlo N-Particle transport code was utilized to determine the detector's spatial efficiency. A dime sized source with the established emission spectra was placed into the center of a rectangle of soil. This rectangle, including the source, was used as a lattice element to fill the region of space below the detector. The simulation was designed to track the energy distribution of pulses created in the sodium iodide crystal using the F8 tally, and was setup to isolate each source; thereby correlating detector response to source position. The simulation output required post-processing to determine conversion factors from counts per second to disintegrations per second. Post-processing accounted for subtraction of the Compton continuum, and average the data over the scanning pattern to obtain the expected spatial sensitivity during a characterization study. This was performed for the following photo-peaks: 609 keV (Bi-214), and 1764 keV (Bi-214). For calculation of the scan minimum detectable concentration, the surveyor efficiency coefficient from NUREG 1507 was supplanted in favor of a scan speed deviation factor. This was used because the detector utilizes computerized data collection (spatial and spectral). This report finds that, for two shielded 10.16 cm x 10.16 cm x 40.64 cm sodium-iodide crystals with integrated multi-channel analyzers, the minimum detectable concentration in surface soils is less than 1 pCi per gram Ra-226 for a 1 second survey measurement. These results assume a volumetric source with a maximum depth of 15.24 cm and a minimum decay time of 21 days. Verification of the system's capabilities was performed by sending soil samples from scanned areas, and locations where in-situ measurements were performed, to an independent laboratory for analysis. The laboratory results demonstrated that scanned area was principally comprised of soil containing less than 1 pCi per gram Ra-226. This borders the system MDC and it underestimated the concentration of radium in the soil. The probable causes of this is escape of radon gas from the surface soil. Radon is the daughter of radium, and has a high mobility in soil. If the gas escapes, then the progeny used as the detection proxy will be present in lower than expected concentrations. Radionuclide identification using high volume sodium-iodide crystals in low level characterization scans is demonstrated to be possible. Further system refinement will focus on developing a methodology to estimate a radon escape factor for site specific use. Accounting for radon escape in the top 15.24 cm of soil will allow the system to better estimate Ra-226 concentrations in soil. This is a significant advancement over the gross counting statistics used in most studies to satisfy MARSSIM survey requirements, because it will increase stakeholder confidence. (authors)

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

Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometer at EAST are studied for future data interpretation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Simulation of fission gas diffusion in UO{sub 2} by vacancy clusters

The low solubility of fission gas in nuclear fuel leads to inter- and intra-granular bubbles. These bubbles affect performance of the fuel rod and increase the fuel temperature through the increase of porosity in the fuel pellets and decreased thermal conductivity across the pellet-clad gap as gas is released into the plenum. The nucleation, growth, and inter-connectivity of the fission gas bubbles ultimately depend on the diffusivity of gas atoms through the fuel matrix. Using mechanisms and data previously established from density functional theory (DFT) and empirical potential calculations, diffusion and reaction models have been derived for xenon atoms, uranium atoms, and vacancies in the substitutional and interstitial locations. Although similar to previous models, the diffusion and reaction rates have been re-derived in a more robust manner. Previous studies have under-predicted the experimentally observed diffusivity of xenon atoms in UO{sub 2}, however it has been previously shown that by including xenon atoms bound to two vacancies, the calculated diffusivity approaches the measured data. The derived model for species diffusion and inter-species reactions has been implemented in the MARMOT phase-field code in a modular structure so as to capture the enhanced diffusion of xenon due to bound gas + vacancy-cluster movement. The MARMOT code allows for the diffusion, reaction, and source rates to be solved simultaneously, resulting in steady state or transient concentrations of each species, and ultimately the diffusion behavior of xenon atoms. By including the diffusion of clusters of different combinations of xenon and vacancies, the calculated diffusivity more closely captures the measured diffusivity of xenon when using preliminary parameters. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Thermal hydraulic model of the molten salt reactor experiment with the NEAMS system analysis module

System analysis codes have a long history of providing best-estimate and conservative safety analysis for both light water and advanced reactor technologies, including molten salt reactors. As interest continues to expand with advanced reactor concepts, system analysis codes will need revisions to accommodate the behavior of these technologies. Legacy system analysis codes will need to be updated to the latest numerical techniques to shorten execution time and increase the accuracy of results. One example of a modern system analysis code that already encompasses these characteristics is the System Analysis Module (SAM). One key objective of this paper was to review available information for system code modeling of the Molten Salt Reactor Experiment (MSRE) from sources in the open literature and collect the information from these open sources in one place for the first time. This supports the potential objective of developing an open specification for system code analysis for MSRE steady state and transients with and without reactor kinetics. Data from actual MSRE tests will serve as the basis for code-to-code comparison exercises, including the MSRE zero power physics tests, the fuel pump start-up and coast down tests, and the natural circulation transient. The objective is to produce a code-to-code benchmark with a standardized set of comparison problems, recognizing the limitations of the original data. To demonstrate an initial application of this objective and the usefulness of compiling this open data, two Molten Salt Reactor Experiment (MSRE)-related models were developed to evaluate SAM for liquid fueled molten salt reactors. One model was the SAM MSRE hydraulic mockup, which provided experimental data for pressure drop measurements. The second model was the complete MSRE primary loop. The MSRE primary loop model incorporated a fluoride salt fuel/coolant with heat transfer in both the core and heat exchanger. For both the hydraulic mockup and MSRE primary loop models, a holistic 1-D system description was built using open documentation, an open description that can be readily modified and applied for any system analysis code. SAM results for the pressure drop of the hydraulic mockup model were within 6% with measurements. Coolant temperatures for the primary loop model matched the expected axial change in temperature from historical calculations. Using alternative coolant properties obtained from the literature, corresponding to salts with different actinide contents, returned similar trends in core temperature profiles. A thermal hydraulic demonstration of a loss-of-flow transient showed the importance of coupling SAM thermal hydraulic analysis to neutronics. Here, this coupling is essential for simulating MSR transients with system analysis codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Accelerated coupled Monte Carlo-Thermal hydraulic calculations using a hybrid GTF-diffusion-based prediction block: first results

Accurate predictions of spatial power and temperature distributions require the coupling of a neutron transport solver with a thermal-hydraulic (TH) feedback. Nowadays, Monte Carlo (MC) codes are widely coupled to TH solvers, typically via a Picard iteration (PI) method, due to the higher fidelity that such frameworks can produce. To speed up a PI, a prediction step can produce an improved initial guess for a source distribution and feed it to the MC code. Recent work investigated a prediction step that uses generalized transfer functions (GTFs) to predict the macroscopic cross sections' variations following a perturbation in TH properties, such as coolant density. The previous method also relied on first order perturbation (FOP) theory to predict perturbed power profiles, rather than using an expensive MC iterate. The implemented FOP method relied on generating a fission matrix from which the forward and adjoint Eigenmodes were extracted and later used to for power calculations. The generation of the fission matrix can introduce a significant computational overhead, therefore undermining the performance of the proposed hybrid technique when applied to high-dimensional problems, e.g., full core calculations. This work attempts to improve the GTF-FOP prediction step by replacing the FOP solver with a nodal diffusion solver, thus eliminating the need to calculate a fission matrix. The GTF-diffusion step was tested for various moderator density perturbations. In each case, the predicted power distribution showed good agreement with the reference case. The latter is attributed to the generally good prediction of most spatially distributed macroscopic cross sections, except the transport cross section, which will become the focus of future work. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Accelerated Coupled Monte Carlo-Thermal Hydraulic Calculations using a Hybrid GTF-Diffusion-based Prediction Block: First Results

Accurate predictions of spatial power and temperature distributions require the coupling of a neutron transport solver with a thermal-hydraulic (TH) feedback. Nowadays, Monte Carlo (MC) codes are widely coupled to TH solvers, typically via a Picard iteration (PI) method, due to the higher fidelity that such frameworks can produce. To speed up a PI, a prediction step can produce an improved initial guess for a source distribution and feed it to the MC code. Recent work [1, 2] investigated a prediction step that uses generalized transfer functions (GTFs) to predict the macroscopic cross sections’ variations following a perturbation in TH properties, such as coolant density. The previous method also relied on first order perturbation (FOP) theory to predict perturbed power profiles, rather than using an expensive MC iterate. The implemented FOP method relied on generating a fission matrix from which the forward and adjoint eigenmodes were extracted and later used to for power calculations. The generation of the fission matrix can introduce a significant computational overhead, therefore undermining the performance of the proposed hybrid technique when applied to high-dimensional problems, e.g., full core calculations. This work attempts to improve the GTF-FOP prediction step by replacing the FOP solver with a nodal diffusion solver, thus eliminating the need to calculate a fission matrix. The GTF-diffusion step was tested for various moderator density perturbations. In each case, the predicted power distribution showed good agreement with the reference case. The latter is attributed to the generally good prediction of most spatially distributed macroscopic cross sections, except the transport cross section, which will become the focus of future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

ADAPTIVE SURROGATES WITHIN THE RAVEN FRAMEWORK FOR DYNAMIC PROBABILISTIC RISK ASSESSMENT ANALYSIS

The RAVEN code has been under development at the Idaho National Laboratory since 2012. Its main goal is to create a multi-purpose platform for the deploying of all the capabilities needed for Probabilistic Risk Assessment (PRA), Uncertainty Quantification (UQ), data mining analysis and optimization studies. RAVEN has demonstrated a good level of maturity in terms of state-of-art and advanced analysis methodologies. In the past year the RAVEN code has been released as open-source project and is available to download (raven.inl.gov) free of charge. The main subject of this paper is to present a new system within the RAVEN framework to drastically accelerate PRA and UQ analysis using acceleration schemes based on surrogate modeling. This system is based on the heavy usage of surrogate modeling and artificial intelligence algorithms, able to emulate the response of a physical system in closed domains (domains where the surrogate models have been constructed from). The aim of this paper is to show to the community how the employment of surrogate modeling techniques can be a viable solution for the current and future challenges in terms of PRA, UQ and Safety Margins characterizations. In addition, in this manuscript it is presented the construction of a RAVEN new entity named “HybridModel” that is aimed to switch between a high-fidelity code and its linked surrogate, accelerating the PRA and UQ processes without largely reducing the accuracy of the results of the analyses. All these new capabilities will be shown with an application example about a Station Black Out scenario for a prototypical Pressurized Water Reactor (modeled with RELAP5-3D).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Model Data for the Mesh Convergence Study Demonstrating Benefits of Mixed-polyhedral Mesh in Integrated Hydrology Simulations

This archived model data is related to a study introducing a unique method that employs a stream-aligned mixed-polyhedral mesh to effectively and accurately represent river valleys, stream corridors, and narrow engineered channels in integrated hydrology simulations. The study finds that utilizing stream-aligned mixed-polyhedral meshes in integrated hydrology simulations achieves accuracy on par with a finely refined TIN-based mesh while markedly diminishing computational costs. This archive contains scripts and data files needed to generate the ATS model input, including mesh and ATS input files, for all mesh scenarios using the Watershed Workflow package. Additionally, this archive also provides key outputs from the model simulations that are used in the analysis and post-processing scripts to reproduce figures in the manuscript. The Watershed Workflow package is implemented in Python3. The Jupyter notebooks can be executed through multiple open-source tools, for example, Anaconda Jupyter Lab, VS Studio Code, etc. Other data files include CSV and HDF5 files, which can be read through Python scripts. The input files for the ATS model, open-source integrated hydrology, and transport model are in XML format and can be edited in any commonly used text editors.

54 ENVIRONMENTAL SCIENCES↗