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At least 1,315 records · Page 73

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NASA NTRS2011

Computing Temperatures in Optically Thick Protoplanetary Disks

We worked with a Monte Carlo radiative transfer code to simulate the transfer of energy through protoplanetary disks, where planet formation occurs. The code tracks photons from the star into the disk, through scattering, absorption and re-emission, until they escape to infinity. High optical depths in the disk interior dominate the computation time because it takes the photon packet many interactions to get out of the region. High optical depths also receive few photons and therefore do not have well-estimated temperatures. We applied a modified random walk (MRW) approximation for treating high optical depths and to speed up the Monte Carlo calculations. The MRW is implemented by calculating the average number of interactions the photon packet will undergo in diffusing within a single cell of the spatial grid and then updating the packet position, packet frequencies, and local radiation absorption rate appropriately. The MRW approximation was then tested for accuracy and speed compared to the original code. We determined that MRW provides accurate answers to Monte Carlo Radiative transfer simulations. The speed gained from using MRW is shown to be proportional to the disk mass.

radiative transfer↗

DOE OSTI2016

Benchmarking PARTISN with Analog Monte Carlo: Moments of the Neutron Number and the Cumulative Fission Number Probability Distributions

The purpose of this report is to provide the reader with an understanding of how a Monte Carlo neutron transport code was written, developed, and evolved to calculate the probability distribution functions (PDFs) and their moments for the neutron number at a final time as well as the cumulative fission number, along with introducing several basic Monte Carlo concepts.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2014

Coupling External Radiation Transport Code Results to the GADRAS Detector Response Function

Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex, scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

NASA NTRS1999

Spaceflight Validation of Hzetrn Code

HZETRN is being developed as a fast deterministic radiation transport code applicable to neutrons, protons, and multiply charged ions in the space environment. It was recently applied to 50 hours of IMP8 data measured during the August 4, 1972 solar event to map the hourly exposures within the human body under several shield configurations. This calculation required only 18 hours on a VAX 4000 machine. A similar calculation using the Monte Carlo method would have required two years of dedicated computer time. The code has been benchmarked against well documented and tested Monte Carlo proton transport codes with good success. The code will allow important trade studies to be made with relative ease due to the computational speed and will be useful in assessing design alternatives in an integrated system software environment. Since there are no well tested Monte Carlo codes for HZE particles, we have been engaged in flight validation of the HZETRN results. To date we have made comparison with TEPC, CR-39, charge particle telescopes, and Bonner spheres. This broad range of detectors allows us to test a number of functions related to differing physical processes which add to the complicated radiation fields within a spacecraft or the human body, which functions can be calculated by the HZETRN code system. In the present report we will review these results.

Wilson, J. W.↗

DOE OSTI2015

A new cubic phantom for PET/CT dosimetry: Experimental and Monte Carlo characterization

In recent years, positron emission tomography (PET) associated with multidetector computed tomography (MDCT) has become a diagnostic technique widely disseminated to evaluate various malignant tumors and other diseases. However, during PET/CT examinations, the doses of ionizing radiation experienced by the internal organs of patients may be substantial. To study the doses involved in PET/CT procedures, a new cubic phantom of overlapping acrylic plates was developed and characterized. This phantom has a deposit for the placement of the fluorine-18 fluoro-2-deoxy-D-glucose ({sup 18}F-FDG) solution. There are also small holes near the faces for the insertion of optically stimulated luminescence dosimeters (OSLD). The holes for OSLD are positioned at different distances from the {sup 18}F-FDG deposit. The experimental results were obtained in two PET/CT devices operating with different parameters. Differences in the absorbed doses were observed in OSLD measurements due to the non-orthogonal positioning of the detectors inside the phantom. This phantom was also evaluated using Monte Carlo simulations, with the MCNPX code. The phantom and the geometrical characteristics of the equipment were carefully modeled in the MCNPX code, in order to develop a new methodology form comparison of experimental and simulated results, as well as to allow the characterization of PET/CT equipments in Monte Carlo simulations. All results showed good agreement, proving that this new phantom may be applied for these experiments. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

NASA NTRS2010

Study of Plume Impingement Effects in the Lunar Lander Environment

Plume impingement effects from the descent and ascent engine firings of the Lunar Lander were analyzed in support of the Lunar Architecture Team under the Constellation Program. The descent stage analysis was performed to obtain shear and pressure forces on the lunar surface as well as velocity and density profiles in the flow field in an effort to understand lunar soil erosion and ejected soil impact damage which was analyzed as part of a separate study. A CFD/DSMC decoupled methodology was used with the Bird continuum breakdown parameter to distinguish the continuum flow from the rarefied flow. The ascent stage analysis was performed to ascertain the forces and moments acting on the Lunar Lander Ascent Module due to the firing of the main engine on take-off. The Reacting and Multiphase Program (RAMP) method of characteristics (MOC) code was used to model the continuum region of the nozzle plume, and the Direct Simulation Monte Carlo (DSMC) Analysis Code (DAC) was used to model the impingement results in the rarefied region. The ascent module (AM) was analyzed for various pitch and yaw rotations and for various heights in relation to the descent module (DM). For the ascent stage analysis, the plume inflow boundary was located near the nozzle exit plane in a region where the flow number density was large enough to make the DSMC solution computationally expensive. Therefore, a scaling coefficient was used to make the DSMC solution more computationally manageable. An analysis of the effectiveness of this scaling technique was performed by investigating various scaling parameters for a single height and rotation of the AM. Because the inflow boundary was near the nozzle exit plane, another analysis was performed investigating three different inflow contours to determine the effects of the flow expansion around the nozzle lip on the final plume impingement results.

Marichalar, Jeremiah↗

DOE OSTI2017

Analysis of the VERA Core Physics Benchmark Problems using cosRMC Code

In reactor physics, two kinds of numerical methods are generally employed to solve the neutron transport equations, one of which is the deterministic method and the other one is the Monte Carlo method. Monte Carlo is widely used to provide reference solution because of its ability to deal with continuous energy cross sections and complex geometry. Just like any statistical process, the Monte Carlo method requires repetition to achieve a small relative uncertainty, and therefore may inevitably have larger simulation times than deterministic method. Fortunately, the great advancements of computer technology make it possible to apply Monte Carlo method in reactor core analysis. To promote nuclear power plant system design and simulation software self-reliance development, a Monte Carlo neutron/photon transport code cosRMC has been developed jointly by State Nuclear Power Software Development Center (SNPSDC) and Tsinghua University. The cosRMC is aimed at achieving full core calculations and analysis with high fidelity and efficiency by means of advanced methodologies and algorithms as well as high performance computing techniques. The cosRMC has many functions as criticality calculation, fixed-source calculation, burnup calculation, etc. In order to ensure its reliability, several types of verification and validation (V and V) work have been and are being carried out for cosRMC. The simulation results of cosRMC based on Problem 1 to 3 of VERA Core Physics Benchmark showed its good capability of 2D lattice and 3D assembly modeling. In this paper, some progressive verification results of multi-assembly and full core problems are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

General Relativistic Implicit Monte Carlo Radiation-hydrodynamics

Abstract We report on a new capability added to our general relativistic radiation-magnetohydrodynamics code, Cosmos++ : an implicit Monte Carlo (IMC) treatment for radiation transport. The method is based on a Fleck-type implicit discretization of the radiation-hydrodynamics equations, but generalized for both Newtonian and relativistic regimes. A multiple reference frame approach is used to geodesically transport photon packets (and solve the hydrodynamics equations) in the coordinate frame, while radiation–matter interactions are handled either in the fluid or electron frames then communicated via Lorentz boosts and orthonormal tetrad bases attached to the fluid. We describe a method for constructing estimators of radiation moments using path-weighting that generalizes to arbitrary coordinate systems in flat or curved spacetime. Absorption, emission, scattering, and relativistic Comptonization are among the matter interactions considered in this report. We discuss our formulations and numerical methods, and validate our models against a suite of radiation and coupled radiation-hydrodynamics test problems in both flat and curved spacetimes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

NASA NTRS2024

A Computationally Efficient Algorithm for Sampling the Rudd Differential Cross Section

Monte Carlo radiation transport codes such as RITRACKS or Geant4 are used to simulate the interaction of ions with matter. These codes rely on sampling algorithms to determine interactions and various physical properties of particles involved in the simulations. It is crucial to develop efficient sampling algorithms since Monte Carlo radiation transport simulations can be time consuming. This work presents an efficient sampling algorithm to determine the energy of secondary electrons following ion-water interactions. The applicability and intended use of the algorithm are discussed in detail, and it is shown that the new algorithm is up to 6X10 4 times faster than the method currently used in Geant4-DNA.

Floriane Poignant↗

DOE OSTI2015

Nuclear signal simulation applied to gas ionizing chambers

Particle transport codes used in detector simulation allow the calculation of the energy deposited by charged particles produced following an interaction. The pulses temporal shaping is more and more used in nuclear measurement into pulse shape analysis techniques. A model is proposed in this paper to simulate the pulse temporal shaping and the associated noise level thanks to the output track file PTRAC provides by Monte-Carlo particle transport codes. The model has been dedicated to ion chambers and more especially for High Pressure Xenon chambers HPXe where the pulse shape analysis can resolve some issues regarding with this technology as the ballistic deficit phenomenon. The model is fully described and an example is presented as a validation of such full detector simulation. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2012

NUMERICAL MODELING OF MULTI-WAVELENGTH SPECTRA OF M87 CORE EMISSION

Spectral fits to M87 core data from radio to hard X-ray are generated via a specially selected software suite, comprised of the High-Accuracy Relativistic Magnetohydrodynamics GRMHD accretion disk model and a two-dimensional Monte Carlo radiation transport code. By determining appropriate parameter changes necessary to fit X-ray-quiescent and flaring behavior of M87's core, we assess the reasonableness of various flaring mechanisms. This shows that an accretion disk model of M87's core out to 28 GM/c{sup 2} can describe the inner emissions. High spin rates show GRMHD-driven polar outflow generation, without citing an external jet model. Our results favor accretion rate changes as the dominant mechanism of X-ray flux and index changes, with variations in density of approximately 20% necessary to scale between the average X-ray spectrum and flaring or quiescent spectra. The best-fit parameters are black hole spin a/M > 0.8 and maximum accretion flow density n {<=} 3 Multiplication-Sign 10{sup 7} cm{sup -3}, equivalent to horizon accretion rates between m-dot = M-dot / M-dot{sub Edd}{approx}2 Multiplication-Sign 10{sup -6} and 1 Multiplication-Sign 10{sup -5} (with M-dot{sub Edd} defined assuming a radiative efficiency {eta} = 0.1). These results demonstrate that the immediate surroundings of M87's core are appropriate to explain observed X-ray variability.

79 ASTRONOMY AND ASTROPHYSICS↗

NASA NTRS2018

Differences in the Gas and Dust Distribution in the Transitional Disk of a Sun-Like Young Star, PDS 70

We present ALMA 0.87 mm continuum, HCO+ J = 4–3 emission line, and CO J = 3–2 emission line data of the disk of material around the young, Sun-like star PDS 70. These data reveal the existence of a possible two-component transitional disk system with a radial dust gap of 0.42 arcsec ± 0.05 arcsec, an azimuthal gap in the HCO+ J = 4–3 moment zero map, as well as two bridge-like features in the gas data. Interestingly these features in the gas disk have no analog in the dust disk making them of particular interest. We modeled the dust disk using the Monte Carlo radiative transfer code HOCHUNK3D using a two-disk component. We ﬁnd that there is a radial gap that extends from 15 to 60 au in all grain sizes, which differs from previous work.

radial gap↗

DOE OSTI2012

Optimization of a coupling scheme between MCNP5 and SUBCHANFLOW for high fidelity modeling of LWR reactors

In order to increase the accuracy and the degree of spatial resolution of core design studies, coupled Three-Dimensional (3D) neutronics (deterministic and Monte Carlo) and 3D thermal hydraulics (CFD and sub-channel) codes are being developed worldwide. In this paper the optimization of a coupling between MCNP5 code and an in-house development thermal-hydraulics code SUBCHANFLOW is presented. Various improvements of the coupling methodology are presented. With the help of novel interpolation tool a consistent methodology for the preparation of thermal scattering data library have been developed, ensuring that inelastic scattering from bound nuclei is treated at the correct moderator temperature. Trough the utilization of a hybrid coupling with discrete energy Monte-Carlo code KENO a methodology for acceleration of the coupled calculation is being demonstrated. In this approach an additional coupling between KENO and SUBCHANFLOW was developed, the converged results of which are used as initial conditions for the MCNP-SUBCHANFLOW coupling. Acceleration of fission source distribution convergence, by sampling fission source distribution from the power distribution obtained by KENO is also demonstrated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

LLNL Mercury Project Trinity Open Science Final Report

The Mercury Monte Carlo particle transport code developed at Lawrence Livermore National Laboratory (LLNL) is used to simulate the transport of radiation through urban environments. These challenging calculations include complicated geometries and require significant computational resources to complete. As a result, a question arises as to the level of convergence of the calculations with Monte Carlo simulation particle count. In the Trinity Open Science calculations, one main focus was to investigate convergence of the relevant simulation quantities with Monte Carlo particle count to assess the current simulation methodology. Both for this application space but also of more general applicability, we also investigated the impact of code algorithms on parallel scaling on the Trinity machine as well as the utilization of the Trinity DataWarp burst buffer technology in Mercury via the LLNL Scalable Checkpoint/Restart (SCR) library.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2012

Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII

Currently the US lacks an end-to-end (i.e., source-to-detector) radiation transport simulation code with predictive capability for the broad range of DHS nuclear material detection applications. For example, gaps in the physics, along with inadequate analysis algorithms, make it difficult for Monte Carlo simulations to provide a comprehensive evaluation, design, and optimization of proposed interrogation systems. With the development and implementation of several key physics and algorithm enhancements, along with needed improvements in evaluated data and benchmark measurements, the MCNP/X Monte Carlo codes will provide designers, operators, and systems analysts with a validated tool for developing state-of-the-art active and passive detection systems. This project is currently in its seventh year (Phase VII). This presentation will review thirty enhancements that have been implemented in MCNPX over the last 3 years and were included in the 2011 release of version 2.7.0. These improvements include 12 physics enhancements, 4 source enhancements, 8 tally enhancements, and 6 other enhancements. Examples and results will be provided for each of these features. The presentation will also discuss the eight enhancements that will be migrated into MCNP6 over the upcoming year.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2013

Comparison of SERPENT and CASMO-5M for pressurized water reactors models

The objective of this work is to perform a preliminary assessment of the capability of SERPENT to generate cross sections for a PWR Beginning-of-Life (BOL) isothermal mini-core by comparing a SERPENT/PARCS solution with the results obtained using a CASMO-5M/PARCS approach. The PARCS code is used instead of the usual SIMULATE-3 to analyze the Swiss Reactors, because interfaces with PARCS already exist to obtain neutronic data from SERPENT. For the PWR configurations, the differences between CASMO-5M and SERPENT solutions are within 200 pcm at the assembly level and thus rather small when considering the deterministic transport method (energy/angular/space discretization) in CASMO-5M versus the stochastic treatment of SERPENT, the statistical uncertainties in the Monte-Carlo approach as well as the eventual differences in nuclear data used by both codes. At the 2D mini-core level, no major difference is observed when comparing PARCS run with CASMO-5M versus SERPENT cross sections. For the generation of kinetic parameters, non trivial differences are observed due both to the methods and the data used. For the relatively limited number of configurations considered, it is hard to make any definitive statement on the benefits of using Monte Carlo codes in terms of nuclear data generation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

SU-E-J-08: Comparison of Unintended Radiation Doses to Organs at Risk Resulting From the Out-Of-Field Therapeutic Beams and From Image-Guidance X-Ray Procedures

Purpose: The unintended radiation dose to organs at risk (OAR) can be contributed from imaging guidance procedures as well as from leakage and scatter of therapeutic beams. This study compares the imaging dose with the unintended out-of-field therapeutic dose to patient sensitive organs. Methods: The Monte Carlo EGSnrc user codes, BEAMnrc and DOSXYZnrc, were used to simulate kV X-ray sources from imaging devices as well as the therapeutic IMRT/VMAT beams and to calculate doses to target and OARs on patient treatment planning CT images. The accuracy of the Monte Carlo simulations was benchmarked against measurements in phantoms. The dose-volume histogram was utilized in analyzing the patient organ doses. Results: The dose resulting from Standard Head kV-CBCT scans to bone and soft tissues ranges from 0.7 to 1.1 cGy and from 0.03 to 0.3 cGy, respectively. The dose resulting from Thorax scans on the chest to bone and soft tissues ranges from 1.1 to 1.8 cGy and from 0.3 to 0.6 cGy, respectively. The dose resulting from Pelvis scans on the abdomen to bone and soft tissues range from 3.2 to 4.2 cGy and from 1.2 to 2.2 cGy, respectively. The out-of-field doses to OAR are sensitive to the distance between the treated target and the OAR. For a typical Head-and-Neck IMRT/VMAT treatment the out-of-field doses to eyes are 1–3% of the target dose, or 2–6 cGy per fraction. Conclusion: The imaging doses to OAR are predictable based on the imaging protocols used when OARs are within the imaged volume and can be estimated and accounted for by using tabulated values. The unintended out-of-field doses are proportional to the target dose, strongly depend on the distance between the treated target and OAR, and are generally higher comparing to the imaging dose. This work was partially supported by Varian research grant VUMC40590.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2012

Explicit temperature treatment in Monte Carlo neutron tracking routines - First results

This article discusses the preliminary implementation of the new explicit temperature treatment method to the development version Monte Carlo reactor physics code Serpent 2 and presents the first practical results calculated using the method. The explicit temperature treatment method, as introduced in [1], is a stochastic method for taking the effect of thermal motion into account on-the-fly in a Monte Carlo neutron transport calculation. The method is based on explicit treatment of the motion of target nuclei at collision sites and requires cross sections at 0 K temperature only, regardless of the number of temperatures in the problem geometry. The method includes a novel capability of modelling continuous temperature distributions. Test calculations are performed for two test cases, a PWR pin-cell and a HTGR system. The resulting k{sub eff} and flux spectra are compared to a reference solution calculated using Serpent 1.1.16 with Doppler-broadening rejection correction [2]. The results are in very good agreement with the reference and also the increase in calculation time due to the new method is on acceptable level although not fully insignificant. On the basis of the current study, the explicit treatment method can be considered feasible for practical calculations. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗