Search NASA⌕ Search

SEARCH · Search NASA

Results for “Deterministic particle method”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

On the use of reverse Brownian motion to accelerate hybrid simulations

Multiscale and multiphysics simulations are two rapidly developing fields of scientific computing. Efficient coupling of continuum (deterministic or stochastic) constitutive solvers with their discrete (stochastic, particle-based) counterparts is a common challenge in both kinds of simulations. We focus on interfacial, tightly coupled simulations of diffusion that combine continuum and particle-based solvers. The latter employs the reverse Brownian motion (rBm), a Monte Carlo approach that allows one to enforce inhomogeneous Dirichlet, Neumann, or Robin boundary conditions and is trivially parallelizable. We discuss numerical approaches for improving the accuracy of rBm in the presence of inhomogeneous Neumann boundary conditions and alternative strategies for coupling the rBm solver with its continuum counterpart. Numerical experiments are used to investigate the convergence, stability, and computational efficiency of the proposed hybrid algorithm.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2022

Photon detector response function methodology using MCNP and shift hybrid radiation transport code for wide-area contamination assay applications

Here, radiation transport modeling using the Monte Carlo N-Particle (MCNP) radiation transport code and Monte Carlo code, Shift, were employed to model detector responses for a variety of wide-area photon contamination scenarios. In this study, 2" × 2" and 3" × 3" cylindrical NaI(Tl) scintillation detector configurations at source detector-distances of 0.5 cm, 1 cm, 2.54 cm, 10 cm, and 30 cm were modeled. Media of soil, concrete, and steel were evaluated for contamination depths ranging from surface to a depth of an infinite thickness in each medium for photon energies ranging from 20 keV to 3 MeV, which correspond to the energies that current detectors can discern. Monoenergetic photon surface contamination detector responses for each of the media, source–detector distances, and detectors were estimated using MCNP v6.2. Shift was harnessed for improved variance reduction of particle transport in highly attenuating media to obtain average cell fluxes in the two MCNP NaI(Tl) scintillation detector configurations. Average cell flux values in Shift were coupled with detector responses from MCNP to convert average cell flux in a void to energy distribution of pulses in the NaI(Tl) scintillation detector crystal of interest. An optimized detector response function methodology was developed by coupling the MCNP radiation transport method with the Consistent Adjoint Driven Importance Sampling (CADIS) hybrid radiation transport method built into Shift to significantly decrease the runtime of thousands of MCNP pulse height simulations. The methodology may be utilized to quickly and accurately facilitate the assessment of a broad range of wide-area environmental contamination assay and decommissioning cleanup applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

E-PVT: enhanced position-velocity-time scheduler for computer-controlled optical finishing with comprehensive considerations of dynamics constraints, continuity and efficiency

Deterministic computer-controlled optical finishing is an essential approach for achieving high-quality optical surfaces. Its determinism and convergence rely heavily on precise and smooth motion control to guide the machine tool over an optical surface to correct residual errors. One widely supported and smooth motion control model is position-velocity-time (PVT), which employs piecewise cubic polynomials to describe positions. Our prior research introduced a PVT-based velocity scheduling method, demonstrating sub-nanometer level convergence in ion beam figuring (IBF) processes. However, three challenges remained. Firstly, this method relies on quadratic programming, resulting in computational intensiveness for dense tool paths. Secondly, the dynamics constraints and velocity and acceleration continuities are not comprehensively considered, limiting the full potential of PVT-based control. Thirdly, no compensation mechanism existed when dynamics constraints are exceeded. In this study, in response to these challenges, we proposed the Enhanced PVT (E-PVT) method, reducing the time complexity from O ( n 3 ) to O ( n ) while fully addressing dynamics constraints and continuities. A novel compensation method utilizing particle swarm optimization was proposed to address situations where dynamics constraints might be exceeded while maintaining the overall processing efficiency. Validation through simulation and experimentation confirmed the improved performance of E-PVT.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Concurrent Monte Carlo transport and fluence optimization with fluence adjusting scalable transport Monte Carlo

Purpose: The future of radiation therapy will require advanced inverse planning solutions to support single-arc, multiple-arc, and “4π” delivery modes, which present unique challenges in finding an optimal treatment plan over a vast search space, while still preserving dosimetric accuracy. The successful clinical implementation of such methods would benefit from Monte Carlo (MC) based dose calculation methods, which can offer improvements in dosimetric accuracy when compared to deterministic methods. The standard method for MC based treatment planning optimization leverages the accuracy of the MC dose calculation and efficiency of well-developed optimization methods, by precalculating the fluence to dose relationship within a patient with MC methods and subsequently optimizing the fluence weights. However, the sequential nature of this implementation is computationally time consuming and memory intensive. Methods to reduce the overhead of the MC precalculation have been explored in the past, demonstrating promising reductions of computational time overhead, but with limited impact on the memory overhead due to the sequential nature of the dose calculation and fluence optimization. The authors propose an entirely new form of “concurrent” Monte Carlo treat plan optimization: a platform which optimizes the fluence during the dose calculation, reduces wasted computation time being spent on beamlets that weakly contribute to the final dose distribution, and requires only a low memory footprint to function. In this initial investigation, the authors explore the key theoretical and practical considerations of optimizing fluence in such a manner. Methods: The authors present a novel derivation and implementation of a gradient descent algorithm that allows for optimization during MC particle transport, based on highly stochastic information generated through particle transport of very few histories. A gradient rescaling and renormalization algorithm, and the concept of momentum from stochastic gradient descent were used to address obstacles unique to performing gradient descent fluence optimization during MC particle transport. The authors have applied their method to two simple geometrical phantoms, and one clinical patient geometry to examine the capability of this platform to generate conformal plans as well as assess its computational scaling and efficiency, respectively. Results: The authors obtain a reduction of at least 50% in total histories transported in their investigation compared to a theoretical unweighted beamlet calculation and subsequent fluence optimization method, and observe a roughly fixed optimization time overhead consisting of ∼10% of the total computation time in all cases. Finally, the authors demonstrate a negligible increase in memory overhead of ∼7–8 MB to allow for optimization of a clinical patient geometry surrounded by 36 beams using their platform. Conclusions: This study demonstrates a fluence optimization approach, which could significantly improve the development of next generation radiation therapy solutions while incurring minimal additional computational overhead.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Extension of the Hauser-Feshbach fission fragment decay model to multichance fission

The Hauser-Feshbach fission fragment decay model, HF 3 D, which calculates the statistical decay of fission fragments, has been expanded to include multichance fission, up to neutron incident energies of 20 MeV. The deterministic decay takes as input prescission quantities—fission probabilities and the average energy causing fission—and postscission quantities—yields in mass, charge, total kinetic energy, spin, and parity. From these fission fragment initial conditions, the full decay is followed through both prompt and delayed particle emissions, allowing for the calculation of prompt neutron and γ properties, such as multiplicity and energy distributions, both independent and cumulative fission yields, and delayed neutron observables. Here, we describe the implementation of multichance fission in the HF 3 D model, and show an example of prompt and delayed quantities beyond first-chance fission, using the example of neutron-induced fission on 235 U. This expansion represents significant progress in consistently modeling the emission of prompt and delayed particles from fissile systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Sensitivity Analysis of Particle-In-Cell Modeling Parameters in Settling Bed, Bubbling Fluidized Bed and Circulating Fluidized Bed

The objective of the work presented is to perform a preliminary sensitivity analysis of particle-in-cell (PIC) model parameters when applied to settling bed, bubbling fluidized bed, and circulating fluidized bed simulations. These examples correspond to widely different flow conditions commonly seen in chemical engineering applications. Simulations were performed using the PIC method in the open-source software Multiphase Flow with Interphase eXchanges (MFiX) developed by the National Energy Technology Laboratory (NETL). As part of the non-intrusive uncertainty quantification (UQ) analysis, simulation campaigns were generated using Nodeworks. Sampling locations or settings for PIC model parameters were determined using the Latin Hypercube method. Response surfaces were created using radial basis functions (RBF), and Sobol’ indices were estimated to quantify the influence of model parameters on the quantities of interest (QoI). This study marks a first step towards systematically determining optimal ranges for model parameters used in MFiX-PIC. Based on limited experience, it is expected that these values would depend strongly on flow conditions. Given the complexity of the multiphase flow systems under analysis, a non-intrusive UQ based approach is used to identify the most influential parameters in each case. This prior knowledge will help in proposing an effective design of experiments (DoE) and determine optimal parameters through techniques such as deterministic or Bayesian calibration, which will be pursued in the future.

42 ENGINEERING↗
DOE OSTI2020

Methods and apparatus for making magnetic skyrmions

A system and a method for the deterministic generation of magnetic skyrmions includes a magnetic strip configured to store and transport skyrmions. The magnetic strip includes one or more spatial inhomogeneities configured to generate a skyrmion at known locations when excited by a current pulse. A current pulse generator is used to inject current pulses into the magnetic strip via contact pads electrically coupled to both the current pulse generator and the magnetic strip. The system also includes a magnetic field source to apply an out-of-plane magnetic field across the magnetic strip to facilitate generation of skyrmions. Skyrmions can be generated by applying an out-of-plane magnetic field to the magnetic strip and injecting a current pulse with sufficient current density towards the spatial inhomogeneities. Once a skyrmion is generated, another current pulse with sufficient current density can be injected to move the skyrmion.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

IAEA CRP on HTGR Uncertainties in Modeling: Assessment of Phase I Lattice to Core Model Uncertainties

Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented by the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Some Notes on Neutron Up-Scattering and the Doppler-Broadening of High-Z Scattering Resonances

When neutrons are scattered by target nuclei at elevated temperatures, it is entirely possible that the neutron will actually gain energy (i.e., up-scatter) from the interaction. This phenomenon is in addition to the more usual case of the neutron losing energy (i.e., down-scatter). Furthermore, the motion of the target nuclei can also cause extended neutron down-scattering, i.e., the neutrons can and do scatter to energies lower than predicted by the simple asymptotic models. In recent years, more attention has been given to temperature-dependent scattering cross sections for materials in neutron multiplying systems. This has led to the inclusion of neutron up-scatter in deterministic codes like Partisn and to free gas scattering models for material temperature effects in Monte Carlo codes like MCNP and cross section processing codes like NJOY. The free gas scattering models have the effect of Doppler Broadening the scattering cross section output spectra in energy and angle. The current state of Doppler-Broadening numerical techniques used at Los Alamos for scattering resonances will be reviewed, and suggestions will be made for further developments. The focus will be on the free gas scattering models currently in use and the development of new models to include high-Z resonance scattering effects. These models change the neutron up-scattering behavior.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

WE-FG-BRA-10: Radiodosimetry of a Novel Alpha Particle Therapy Targeted to Uveal Melanoma: Absorbed Dose to Organs in Mice

Purpose: The melanocortin-1 receptor (MC1R) is expressed in 94% of uveal melanomas and is described as an ideal target for this untreatable disease. MC1RL is a high affinity MC1R specific peptidomimetic ligand that can serve as a scaffold for therapeutic conjugates such as alpha particle emitting isotopes. The purpose of this study was to assess normal tissue distribution and risk as a result of using the DOTA chelator conjugated to MC1RL to deliver {sup 225}Ac: MC1RL-DOTA-{sup 225}Ac. Methods: 17 non-tumor bearing BALB/c mice were intravenously injected with the novel MC1RL-DOTA-{sup 225}Ac radiopharmaceutical with an average initial administered activity of 2.5 µCi. After the injection, three groups of animals (6, 6, and 5 per group) were euthanized at 24, 48, and 96 hour time points. A total of 11 organs of interest were harvested at each time point including kidneys and liver. Since the emitted alpha particles from {sup 225}Ac and its daughter products are not easy to detect directly, the isomeric gamma spectra were measured instead in the tissue samples using a modified Atomlab™ Gamma Counter (Biodex Medical Systems, Inc) and converted using factors for gamma ray abundance per alpha decay. Dosimetry was performed using measured radioactivity distribution in organs and the generalized internal dosimetry schema of MIRD pamphlet #21. Results: Our calculations have shown that the maximum absorbed dose was delivered to the liver with a total of 47 cGy per 96 hour period. The average dose per kidney was calculated to be 21 cGy. Heart, brain, lung, spleen, skin doses ranged from 0.01 to 1 cGy over the same time period. All animals gained weight over the 110 day decay period and no organ damage was observed by pathology. Conclusion: Based on our results, the risk of using the MC1RL-DOTA-{sup 225}Ac compound is relatively small in terms of deterministic radiation effects. Funding Support: NIH/NCI P50CA168536-03 Skin SPORE; NIH/NCI Phase I SBIR Contract #HHSN261201500067C; Imaging and Technology Center of Excellence at Moffitt. Disclosures and Conflict of Interest: Collaboration with Modulation Therapeutics, Inc.(MTI) and has been partially funded by sub-contracts from MTI via collaboration on a NIH/NCI phase I SBIR contract.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2024

DECOVALEX-2023: Task F1 Final Report

DECOVALEX-2023 Task F is a comparison of models and methods for post-closure performance assessment (PA) of a deep geologic repository for radioactive waste. The general aims of Task F are to build confidence in the models, methods, and software used for PA and to stimulate additional research and development in PA methodologies. The task objectives are to motivate development of PA modelling skills and capabilities, to examine the influence of model choices on calculated repository performance, and to compare the uncertainties introduced by model choices to other sources of uncertainty. Task F involves no actual experiment or site. It is a PA modelling exercise that requires the conceptual development of hypothetical repository designs and geologic settings. Because three of the teams were interested in salt and the rest of the teams were interested in crystalline rock, Task F was split into two branches: Task F1 for crystalline rock and Task F2 for salt. This report is for Task F1, crystalline rock. Teams from seven countries (Canada, Czech Republic, Germany, Korea, Sweden, Taiwan, and United States) participated in Task F1. The teams worked together to define the features, events, and processes of the reference case repository and established a set of performance measures. In addition, they defined a set of benchmark problems designed to test and compare modelling capabilities for fracture flow and transport at different scales. The repository design and benchmark problems are documented in a Task Specification that evolved over time as the group honed the specifications. The benchmark problems verified that each team can aptly model flow and transport in fractured media in 1-, 2-, and 3-dimensions. Two general approaches were used for the 3-dimensional benchmarks: discrete fracture network (DFN) and equivalent continuous porous medium (ECPM). DFN modelling involves explicit meshing of each fracture while ECPM modelling aims to capture the effective porosity and directional permeability of each cell in a space-filling mesh as affected by intersecting fractures. In some models, a combination of the two is used, i.e., DFN for large known fractures and ECPM for the rest of the domain. Transport is solved by using either the advection-dispersion equation or particle tracking. Although some variation is observed among model breakthrough curves in the benchmark problems, there is strong agreement in breakthrough behaviour up to at least the 75 th percentile for all benchmarks. At the 90 th percentile, breakthrough results show larger differences, suggesting several models retain substantially higher fractions of tracer in regions of slower moving water. In addition to the flow and transport benchmarks, several teams completed the source term benchmark, verifying capabilities for modelling radionuclide decay and ingrowth, waste package breach, instant release fractions, fuel matrix degradation rates, and radionuclide solubility limitations. The reference case is conceptualized as a generic spent fuel repository at a depth of 450 m in fractured crystalline rock. The repository has 50 parallel backfilled drifts, each with 50 deposition holes 6 m apart. Each deposition hole contains a 4-PWR waste package and bentonite buffer. The rock domain is 5 km in length, 2 km in width, and 1 km in depth. It has 6 deterministic fractured deformation zones and a multitude of stochastic fractures. Teams generally used the ECPM approach for the entire rock or a hybrid approach in which the deterministic fracture zones are modelled with a DFN and the rest of the rock is modelled by ECPM. Of the reference case problems specified, only the results of the initial reference case problem are compared in this report. The initial problem focuses on transport from the deposition holes to the surface, i.e., it neglects waste package performance. Tracers are released at all waste package locations at time zero and tracked for their releases to the near field and ground surface. The water fluxes calculated at the ground surface entry and exit regions of the domain are similar for all models except for two that have considerably lower fluxes. For tracer transport, large differences are observed among models in the magnitude of tracer transported. Much of the difference appears to be due to how the repository is implemented and hence the different degrees of repository simplification. Models that exclude the drifts, buffer, and backfill from the domain tend to show greater release of tracers and radionuclides from the repository. The initial study presented here indicates that major differences in modelling important processes within the repository (e.g., diffusion through buffer and backfill) can produce broadly different release and transport results, especially when those processes are excluded. Even for the models that included all specified features, events, and processes, the results show significant differences and demonstrate the importance of examining multiple modelling approaches in performance assessment. The differences in results observed in this study are expected to motivate teams to either increase complexity in future versions of the reference case models or to improve methods to account for the effects of simplified features and processes. Either way, future improvements in these models are expected to produce results that more closely agree.

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

Spin–orbit torques in structures with asymmetric dusting layers

Current-induced spin–orbit torques (SOTs) in heavy metal/ferromagnet heterostructures have emerged as an efficient method for magnetization switching with applications in nonvolatile magnetic memory and logic devices. However, experimental realization of SOT switching of perpendicular magnetization requires an additional inversion symmetry breaking, calling for modifications of the conventional SOT heterostructures. In this work, we study SOTs and deterministic switching of perpendicular magnetization by inserting different asymmetric dusting layers at the heavy metal/ferromagnet interface. Similar to the previous works with lateral structural asymmetry, we study the emergence of current-induced perpendicular effective magnetic fields (H z eff ). By examining three different material combinations of heavy metal/dusting layers (W/IrMn, Pt/IrMn, and W/Ta), we shed light on the origins of H z eff ; we show that H z eff is generically created in all the studied asymmetric structures, has a close correlation with the interfacial magnetic anisotropy, and is independent of the signs of spin Hall angles of the materials. Furthermore, we show that the induction of H z eff enables field-free deterministic SOT switching of perpendicular magnetization. Our results can be used in designing SOT heterostructures for practical applications in nonvolatile technologies.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Criticality Calculations with MCNP6 - Practical Lectures

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Validation of a Monte Carlo model for the MLC-Elekta Precise with MCNP6 - Paper 76

Cancer treatment using radiotherapy techniques needs an accurate prediction of the dose delivered to the patient at each position. Treatment planning systems (TPS) allow the optimization of the dose delivered to the tumor volume while minimizing the dose absorbed by the surrounding healthy tissue. Commercial treatment planning systems use empirical deterministic algorithms that do not correctly estimate the dose in the surroundings of heterogeneities. This lack of accuracy is taken into account as bigger security margins in the treatment, reducing the potential effect of the treatment. However, Monte Carlo (MC) simulation techniques have proved to predict the dose distribution more accurately and also representing experimental data better. Nevertheless, the clinical application of these techniques has been limited by the long calculation times they require. As high performance computers decrease in cost MC simulations become more feasible to be used in TPS. The present work improves a simplified model of the MLC-Elekta Precise linear accelerator (linac or unit from now on) previously developed and correctly validated by the group. A multileaf collimator (MLC) has been modelled according to the manufacturer specifications and added to the previous model to allow higher conformity in the studies. The full geometry has been entered to the MCNP6 Monte Carlo code to simulate the transport of particles through the linac. The geometry also includes a cubic water phantom for dosimetry studies and validation with experimental data. The experimental data employed for the dose distribution comparison has been measured at the linac from the linac Hospital Universitari Clinic de Valencia (Spain). Additionally, particles are stored in a phase space surface before their entrance in the water to be analyzed and for performance optimization. Given the results presented, we conclude that this first approach to validate a model of the Elekta Precise linac from the Hospital Cl nic Universitari de Valancia represents well the depth-dose distribution, but fails at representing the dose profiles. It has been shown that they predict too much penumbra, probably caused by the filtration between the leaves. In order to solve this problems, further work is needed to improve the simulated model. This work will mainly focus on the multileaf collimator. Also, the codes developed by the group to analyze phase space files produced by MCNP6 have proven highly useful. As work progress and the codes are debugged, they will be published so that the community can take advantage of them.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Comparing Thick Target Neutron Yield Calculations for Alphas on F-19 using MCNP6 1.1 Beta with TENDL 2012 Libraries

Because of the heritage of MCNPX, the first production release of MCNP6{sup TM} used physics models, instead of ACE (A Compact Evaluated Nuclear Data File) formatted libraries, for simulating the transport of alpha particles at all energies, as MCNPX could not yet read ACE formatted light ion libraries. For non-Coulombic interactions, the physics models have suggested minimums of ∼ 20-200 MeV. The first production release of MCNP6, MCNP6{sup TM} 1.0, was still capable of transporting alphas to low energies; however, the physics models resulted in poor prediction of inelastic reactions at energies below the suggested minimums. This poorer prediction at low energies resulted in orders of magnitude lower neutron yields for events resulting from light ion interactions at low energies. For example, F-19(α,n)Na-22 from U-234 decay in UF{sub 6} produced poor neutron production results. Modeling these types of interactions correctly is vital for signal analysis for nonproliferation safeguards and nuclear security as well as for correctly characterizing source terms for radiation protection and shielding calculations. The TALYS computer code was developed to simulate nuclear reactions involving neutrons, photons, protons, deuterons, tritons, alphas and He-3 particles, in the 1 keV - 200 MeV energy range. TENDL is a nuclear data library, which provides the output of the TALYS nuclear model code system for direct use in applications (i.e. ACE formatted cross sections). Since 2008, TENDL had publicly provided ACE formatted libraries for alpha transport; however, not until the release of MCNP6{sup TM} 1.1 beta, had MCNP6 publicly possessed the capability to read such libraries. At the Radiation Protection and Shielding Division topical meeting in 2014, initial testing of MCNP6 1.1 beta showed that MCNP6 using TENDL 2012 resulted in poor prediction when compared to the compilation of experiment data provided in R. Heaton et al. Further comparisons for PuBe sources, using the MCNP6 1.1 beta spontaneous decay alpha feature coupled with alpha transport, were also given in the winter American Nuclear Society meeting in 2014, where the results showed large discrepancies with measured data. For F-19(α,n)Na-22 interactions, the R. Heaton et al. paper is a conglomeration of several important references; and the data provided by R. Heaton et al. used a numerical fit of the excitation function provided by Norman et al.. Norman et al. has recently published their (α,n) thick target yield data for PbF{sub 2} in tabular form, backing out the thick target yields of F-19(α,n)Na-22 and computing F{sub 2} and UF{sub 6} thick target neutron yields (as supplements to the PbF{sub 2} data). Here we compared measured thick target yields of E. B. Norman et al. to MCNP6 1.1 beta using TENDL 2012 libraries. Deterministically backing out the thick target yield from measured values involves using the stopping power. MCNP uses either the SPAR code or an interpolation of the Spar code and Bethe-Boch formula for computing stopping powers. As a result one may believe that these separate methods will arrive at separate thick target yields independent of the production cross section. In our MCNP calculations, we simulated the alpha until the alpha completely stopped in the medium. The energy loss in the condensed history algorithm is not continuous but discrete and related to the stopping powers. Because the neutron production yield is energy dependent, the energy loss dictates the emission of secondary neutrons; and therefore the difference in stopping powers used may have some slight effect on the yield difference. For this study, we chose to use the default condensed history algorithm setup, while future studies should address tightening steps in the condensed history algorithm to prove whether or not choice of stopping power actually matters. The fact that all 5 evaluated isotopic compositions resulted in different thick target yield percent differences does warrant this type of investigation. A typical user may not be only interested in integral emission, taking 10 s to 100 s of processor hours to simulate, but rather a timely generation of the spectra. Such an application would require biasing the production of secondaries from alpha sources such that millions of secondaries could be produced, emulating a converged energy dependent spectra. Neutron production biasing from alpha sources does not currently exist; however, developing a biasing algorithm for neutron production from alpha sources should be investigated in the future.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Operator learning for predicting multiscale bubble growth dynamics

We report simulating and predicting multiscale problems that couple multiple physics and dynamics across many orders of spatiotemporal scales is a great challenge that has not been investigated systematically by deep neural networks (DNNs). Herein, we develop a framework based on operator regression, the so-called deep operator network (DeepONet), with the long-term objective to simplify multiscale modeling by avoiding the fragile and time-consuming “hand-shaking” interface algorithms for stitching together heterogeneous descriptions of multiscale phenomena. To this end, as a first step, we investigate if a DeepONet can learn the dynamics of different scale regimes, one at the deterministic macroscale and the other at the stochastic microscale regime with inherent thermal fluctuations. Specifically, we test the effectiveness and accuracy of the DeepONet in predicting multirate bubble growth dynamics, which is described by a Rayleigh–Plesset (R–P) equation at the macroscale and modeled as a stochastic nucleation and cavitation process at the microscale by dissipative particle dynamics (DPD). First, we generate data using the R–P equation for multirate bubble growth dynamics caused by randomly time-varying liquid pressures drawn from Gaussian random fields (GRFs). Our results show that properly trained DeepONets can accurately predict the macroscale bubble growth dynamics and can outperform long short-term memory networks. We also demonstrate that the DeepONet can extrapolate accurately outside the input distribution using only very few new measurements. Subsequently, we train the DeepONet with DPD data corresponding to stochastic bubble growth dynamics. Although the DPD data are noisy and we only collect sparse data points on the trajectories, the trained DeepONet model is able to predict accurately the mean bubble dynamics for time-varying GRF pressures. Taken together, our findings demonstrate that DeepONets can be employed to unify the macroscale and microscale models of the multirate bubble growth problem, hence providing new insight into the role of operator regression via DNNs in tackling realistic multiscale problems and in simplifying modeling with heterogeneous descriptions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

SU-E-T-270: Diffusion Synthetic Acceleration for Linear Boltzmann Transport Equation

Purpose: Linear Boltzmann transport equation (LBTE) is as accurate as the Monte Carlo method (MC) for dose calculation in photon/particle therapy (LBTE is a deterministic and Eulerian formulation and MC is a statistical and Lagrangian description). An advantage of LBTE is that numerous acceleration techniques can be utilized for acceleration. This work is to explore the acceleration of LBTE via diffusion synthetic acceleration (DSA). Methods: For simplicity, two-dimensional, steady-state, and within-group LBTE is considered with two angular dimensions and two spatial dimensions. The discrete ordinate method is developed for solving this integro-differential equation. The angular variables are discretized using a level-symmetric quadrature set on the unit sphere. The spatial variables are discretized on the structured grid based on the diamond scheme. The source-iteration method (SI) is used to solve the discretized system.Since SI is slow in optically thick and highly scattering regime. DSA is developed to accelerate SI. The motivation for DSA is that diffusion equation (DE) is a good approximation of LBTE in the above regime. However, DE is much cheaper than LBTE computationally since DE only involves spatial variables. Thus, in each DSA iteration, DSA adds to the SI step a computationally-negligible DE step, i.e., to first solve DE with the SI residual as source term, and then compensate the SI solution with DE solution. Results: DSA was benchmarked and compared with SI. The difference between two methods was within 0.12% which verifies the accuracy of DSA, while DSA demonstrated the great advantage in speed, e.g., the reduction of iteration number to 6% and 4% respectively for cases with 100 and 1,000 scattering-absorption ratio that commonly occur in clinical dose calculation. Conclusion: DSA has been developed as one of many possible means for accelerating the numerical solver of LBTE for dose calculation. The authors were partially supported by the NSFC (#11405105), the 973 Program (#2015CB856000) and the Shanghai Pujiang Talent Program (#14PJ1404500)

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

The FREYA project at VENUS-F - the next step towards MYRRHA

The VENUS-F reactor is a fast, zero-power research reactor. Since 2009 it has used 30% wt. metallic uranium as fuel and solid lead as reflector and coolant simulator. It can also be coupled to a particle accelerator with a neutron producing target and make up a subcritical Accelerator Driven System (ADS). Within the currently ongoing FREYA project, the main tasks being investigated are the validation of online reactivity monitoring of an ADS, the validation of nuclear data and neutronic codes, and the experimental characterization of fast critical and subcritical cores representative for MYRRHA. MYRRHA will be an ADS demonstrator with the option to be operated in critical mode too. So far the only critical core at VENUS-F was assembled using 97 fuel assemblies, each of them consisting of 9 U and 16 Pb rodlets. The core was thoroughly characterized: flux traverses, spectral indices, control rod worth, delayed neutron parameters were measured and compared with Monte Carlo and deterministic calculations. A sensitivity study was also performed focusing on the influence of material impurities, positioning uncertainties, structures and detectors around the core. It was a simple core without any perturbations in the active zone. In the next critical cores that are planned to be investigated in 2015 and 2016, additional elements will be introduced in order to become more representative for MYRRHA. As MYRRHA will use MOX fuel, oxygen will be added into the following core. This will be done by introducing Al{sub 2}O{sub 3} into the fuel assemblies. New fuel assemblies have been already made and are ready to be loaded. Each consists of 13 U, 8 Pb, and 4 Al{sub 2}O{sub 3} rodlets. The new critical core will consist of approximately 40 such fuel assemblies and will be loaded in January 2015. The first results of the characterization measurements will be presented. Additionally, it is planned to investigate coolant void effects. As MYRRHA will be a rather complex system using a number of in-pile sections and heterogeneous elements inside and around the active zone, the next cores will be step by step modified by loading new materials and perturbations with the aim to assemble a MYRRHA mock-up. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗