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At least 1,261 records · Page 70

Testing Different Methods for Estimating Uncertainties on Rietveld Refined Parameters Using SrRietveld

The advent of fast computing allows more computationally expensive, though possibly more accurate, approaches for estimating uncertainties on Rietveld refined parameters. Here we compare three such methods, using two different refinement programs, FullProf and GSAS. This is facilitated by the use of a new Rietveld refinement package, SrRietveld, that provides Python wrappers for FullProf and GSAS. The refined values on the parameters from two different refinement engines match each other very well. The uncertainty estimates determined using the different methods are also consistent, though FullProf and GSAS estimate the parameter uncertainties slightly differently from each other, which is discussed. More importantly, we find that the refined results are very sensitive to the statistical weights used in the least squares equation. Different weights, for example from uncertain or incorrectly propagated random errors on the data, lead to significantly different refined values. This means that uncertainty estimates on refined parameters should be increased when the random errors on the data are not well known, as for example in many cases where area detectors are used, and care should be taken to propagate errors correctly in any data preprocessing such as normalization by a smoothed spectrum. The computationally expensive bootstrap error estimation methods are facilitated by the use of SrRietveld, but are not required in most cases though may be useful in some cases, for example if the random errors on the data are not well known.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

An Integrated Signaling-Encryption Mechanism to Reduce Error Propagation in Wireless Communications: Performance Analyses

Traditional encryption techniques require packet overhead, produce processing time delay, and suffer from severe quality of service deterioration due to fades and interference in wireless channels. These issues reduce the effective transmission data rate (throughput) considerably in wireless communications, where data rate with limited bandwidth is the main constraint. In this paper, performance evaluation analyses are conducted for an integrated signaling-encryption mechanism that is secure and enables improved throughput and probability of bit-error in wireless channels. This mechanism eliminates the drawbacks stated herein by encrypting only a small portion of an entire transmitted frame, while the rest is not subject to traditional encryption but goes through a signaling process (designed transformation) with the plaintext of the portion selected for encryption. We also propose to incorporate error correction coding solely on the small encrypted portion of the data to drastically improve the overall bit-error rate performance while not noticeably increasing the required bit-rate. We focus on validating the signaling-encryption mechanism utilizing Hamming and convolutional error correction coding by conducting an end-to-end system-level simulation-based study. The average probability of bit-error and throughput of the encryption mechanism are evaluated over standard Gaussian and Rayleigh fading-type channels and compared to the ones of the conventional advanced encryption standard (AES).

advanced encryption standard (AES)↗

Quicksilver

Quicksilver is proxy application for the Monte Carlo Transport code, Mercury. The purpose of the Quicksilver mini-app is to create a freely distributable code base that represents the key elements of the Mercury workload by solving a simplified time-dependent neutron transport problem. As an open source code, Quicksilver will be hosted on the CZ git/stash repo alongside other mini-apps in the ACS so-design suite. It will be used by vendors and other partners for co-design activities such as identifying root causes of Mercury performance bottlenecks on future architectures (i.e., memory bandwidth, memory latency, flop intensity) and developing Quicksilver will be used within LLNL as a nimble platform to evaluate programming models, algorithms, etc. that might be useful in Mercury. Quicksilver may also become a replacement for MCB in future procurement benchmarks, as well as a way to engage vendors in the Trinity and Sierra Centers of Excellence as well as other co-design efforts.

Richards, David↗

Target Design Optimization of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States developed and designed a neutron source facility for Kharkov Institute of Physics and Technology (KIPT) of Ukraine. The facility was constructed at Kharkov, Ukraine and its commissioning process has been started. The facility has an electron accelerator driving a subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. The subcritical assembly has WWR-M2 fuel assemblies with U-235 enrichment of 19.7 wt%. The facility will be utilized to perform basic and applied nuclear research, to produce medical isotopes, and to train young nuclear specialists. Solid target design with stacked disks is selected and each target disk is cooled by water from both side. Tungsten or natural uranium is the target material. This paper presents the target design optimization to maximize the neutron yield and the neutron flux level of the facility while satisfying the thermal-hydraulic design criteria. Monte Carlo computer code MCNPX is utilized for the neutron yield analyses with ENDF/B-VII.0 nuclear data libraries, as a function of target thickness using a simplified target model without cladding or coolant channels. It’s found that the neutron yield saturates at target thickness 60 ~ 70 mm for both tungsten and uranium materials. For the final target design with cladding and coolant channels, the total uranium thickness is 56.5 mm, and the total tungsten thickness is only 33.0 mm. The tungsten material has a large absorption cross section for thermal neutrons. If the tungsten thickness is increased, the neutron yield gain is offset by the neutron absorption reaction. The paper presents the design optimization analyses for both target materials.

electron accelerator↗

Improved Moderator Performance Calculations at SNS

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory reached a nominal operating power of 1 MW in September, 2009. Even at ~2/3 of its designed capacity, SNS already becomes the brightest pulsed neutron source in the world, and a premier facility for material and life science research. To provide accurate target performance estimation and to meet future development needs, we had developed an as-built model for the operating target system of SNS. Based on this asbuilt model, the moderator performance at SNS was recalculated taking advantage of the improved physics and tallying method in the Monte Carlo simulation code, MCNPX, and most recent scattering kernels. The neutron flux intensity spectra, emission time distributions of neutron pulses and moderator brightness profiles regarding the SNS moderator performance are presented and discussed. Compared to the previous results, the current calculations show ~10-20% lower in flux with improved emission time distributions of neutron pulses. The difference between the two data sets is analyzed, as well as other factors impacting the moderator performance at SNS. The current calculation results will serve as source terms for the instrument performance simulations at SNS.

Lu, Wei↗

gRASPA

GPU Monte Carlo Simulation Code with a taste of RASPA We present enhancements in Monte Carlo simulation speed and functionality within an open-source code, gRASPA, which uses graphical processing units (GPUs) to achieve significant performance improvements compared to serial, CPU implementations of Monte Carlo. The code supports a wide range of Monte Carlo simulations, including canonical ensemble (NVT), grand canonical, NVT Gibbs, Widom test particle insertions, and continuous-fractional component Monte Carlo. Implementation of grand canonical transition matrix Monte Carlo (GC-TMMC) and a novel feature to allow different moves for the different components of metal-organic framework (MOF) structures exemplify the capabilities of gRASPA for precise free energy calculations and enhanced adsorption studies, respectively. The introduction of a High-Throughput Computing (HTC) mode permits many Monte Carlo simulations on a single GPU device for accelerated materials discovery. The code can incorporate machine learning (ML) potentials. The open-source nature of gRASPA promotes reproducibility and openness in science, and users may add features to the code and optimize it for their own purposes. The code is written in CUDA/C++ and SYCL/C++ to support different GPU vendors. The gRASPA code is publicly available at https://github.com/snurr-group/gRASPA.

Li, Zhao [Purdue/Northwestern/Notre Dame Universit↗

Collision Tracking in OpenMC: Methods and Applications in Neutron Noise, Neutron Imaging, Time-of-Flight, and Multiplicity Counting

We present the development and application of a collision tracking feature within the OpenMC Monte Carlo particle transport code, designed for diverse applications such as neutron spectroscopy, scatter camera system, neutron noise, and multiplicity counting simulations. This feature enables the tracking of individual particle collisions, with potential applications in nuclear nonproliferation, reactor physics, and nuclear security. Additionally, the feature holds potential for the calibration of neutron detectors, specifically in converting light output into energy deposited within the detectors. The implementation consists of a set of filters—such as reaction type, energy, cell, and material—that constrain the set of collisions that are tracked, extensions to the Python API to enable simple input specification, and support for writing either OpenMC’s native HDF5-based format or the Monte Carlo particle list format. This feature was added to the official OpenMC release in version 0.15.3. In this work, the feature will be applied to showcase scenarios such as time-of-flight simulations, scatter-camera imaging for neutron source localization, neutron-noise analysis to extract integral kinetic parameters such as the prompt decay constant α, and multiplicity counting to estimate the mass of special nuclear materials. Ultimately, this feature aims to expand the application scope of open-source Monte Carlo particle transport codes such as OpenMC.

Monte Carlo code↗

Introduction to an MCGIDI Mini-App and Performance Comparisons with XSBench

In high-performance Monte Carlo (MC) radiation transport codes, cross section lookups often account for the majority of computational expens. In response to this observation, a number of nuclear data mini-apps have been developed to profile and optimize the lookup process. A review of these mini-apps can be found in Ref. In contrast to a fully fledged MC code, nuclear data mini-apps do not feature particle tracking; instead, the cross section lookup process is considered in isolation, a simplification that allows for easier implementation and evaluation of various lookup schemes and techniques. GIDI+, as described on its public repository1, is ...a collection of C++ libraries for accessing evaluated and processed nuclear data stored in the Generalized Nuclear Database Structure (GNDS). In addition to reading GNDS files, GIDI+ has functions to sum and collapse multi-group data as needed by deterministic transport codes, and to sample GNDS data as needed by Monte Carlo transport codes. Certain modes of Mercury, a LLNL-developed MC radiation transport code, utilize GIDI+ for cross section lookups. Specifically, Mercury uses the MCGIDI library contained within GIDI+ to perform lookups on GPUs. A mini-app exercising MCGIDI’s cross section lookup capabilities would prove useful for optimization, which in turn could improve the performance of client codes such as Mercury. Such a mini-app could also be used to assess the performance of MCGIDI compared to other nuclear data mini-apps. In this document, we introduce a nuclear data mini-app built with the MCGIDI library. The MCGIDI mini-app (MCGIDI-MA) was developed to closely resemble the methodology present in XSBench, a nuclear data mini-app developed at Argonne National Laboratory. The remainder of this section provides an overview on XSBench. Section 2 describes MCGIDI-MA and its usage. Section 3 contains performance results of MCGIDI-MA on LLNL’s Lassen and Quartz compute platforms, as well as comparisons to XSBench performance where appropriate. In Section 4, we present a summary of our findings.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Monte Carlo Radiation Transport for Astrophysical Transients Powered by Circumstellar Interaction

In this paper, we introduce SuperLite, an open-source Monte Carlo radiation transport code designed to produce synthetic spectra for astrophysical transient phenomena affected by circumstellar interaction. SuperLite utilizes Monte Carlo methods for semi-implicit, semirelativistic radiation transport in high-velocity shocked outflows, employing multigroup structured opacity calculations. The code enables rapid post-processing of hydrodynamic profiles to generate high-quality spectra that can be compared with observations of transient events, including superluminous supernovae, pulsational pair-instability supernovae, and other peculiar transients. We present the methods employed in SuperLite and compare the code's performance to that of other radiative transport codes, such as SuperNu and CMFGEN. We show that SuperLite has successfully passed standard Monte Carlo radiation transport tests and can reproduce spectra of typical supernovae of Type Ia, Type IIP, and Type IIn.

79 ASTRONOMY AND ASTROPHYSICS↗

Simulations of particle acceleration in parallel shocks: Direct comparison between Monte Carlo and one-dimensional hybrid codes

We have made a direct comparison between two different computer simulations of a plane, parallel, collisionless shock including particle acceleration to energies typical of those of diffuse ions observed at the earth bow shock. Despite the fact that the one-dimensional hybrid and Monte Carlo techniques employ entirely different algorithms, they give surprisingly close agreement in the overall shapes of the complete distribution functions for protons as well as heavier ions. Both methods show that energetic ions emerge smoothly from the background thermal plasma with approximately the same relative injection rate and that the fraction of the incoming plasma's energy flux that is converted into downstream enthalpy flux of the accelerated population (i.e., the acceleration efficiency) is similar in the two cases. The fraction of the downstream proton distribution made up of superthermal particles is quite large, with at least 10% of the energy flux going into protons with energies above 10 keV. In addition, an upstream precursor, produced by backstreaming energetic particles, is present in both shocks, although the Monte Carlo precursor is considerably longer than that produced in the hybrid shock. These results offer convincing evidence that, at least in these ways, the two simulations are consistent in their description of parallel shock structure and particle acceleration, and they lay the groundwork for development of shock models employing a combination of both methods.

Ellison, Donald C.↗

Improved modeling of relativistic collisions and collisional ionization in particle-in-cell codes

An improved Monte Carlo collisional scheme modeling both elastic and inelastic interactions has been implemented into the particle-in-cell code CALDER[E. Lefebvre et al., Nucl. Fusion 43, 629 (2003)]. Based on the technique proposed by Nanbu and Yonemura [J. Comput. Phys. 145, 639 (1998)] allowing to handle arbitrarily weighted macro-particles, this binary collision scheme uses a more compact and accurate relativistic formulation than the algorithm recently worked out by Sentoku and Kemp [J. Comput. Phys. 227, 6846 (2008)]. Our scheme is validated through several test cases, demonstrating, in particular, its capability of modeling the electrical resistivity and stopping power of a solid-density plasma over a broad parameter range. A relativistic collisional ionization scheme is developed within the same framework, and tested in several physical scenarios. Finally, our scheme is applied in a set of integrated particle-in-cell simulations of laser-driven fast electron transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Oak Ridge National Laboratory Shutdown Dose Rate Code Suite

The Oak Ridge National Laboratory (ORNL) shutdown dose rate (SDDR) code suite calculates gamma dose rates at a location of interest due to activation of materials. The code suite is based on the rigorous two-step (R2S) method that involves two radiation transport calculations: (1) neutron radiation transport to determine neutron fluxes in the material for which activation calculations will be performed and (2) gamma radiation transport to determine dose rates at a location of interest, due to material activation. The ORNL SDDR code suite is used to determine an importance function that characterizes the neutron’s importance to the final SDDR when the Monte Carlo radiation transport method is used. This report demonstrates how the ORNL SDDR code suite is implemented through the following codes: Monte Carlo N-Particle (MCNP) code Version 5-1.60, Oak Ridge Isotope Generator (ORIGEN), MSX suite of utilities, Neutron Activation Gamma Source Sampler (NAGSS), and Automated Variance Reduction Generator (ADVANTG). For this demonstration, the International Thermonuclear Experimental Reactor (ITER) SDDR benchmark problem is used and the dose rate result is compared to previously published work.

61 RADIATION PROTECTION AND DOSIMETRY↗

SU-E-T-525: Ionization Chamber Perturbation in Flattening Filter Free Beams

Purpose: Changing the characteristic of a photon beam by mechanically removing the flattening filter may impact the dose response of ionization chambers. Thus, perturbation factors of cylindrical ionization chambers in conventional and flattening filter free photon beams were calculated by Monte Carlo simulations. Methods: The EGSnrc/BEAMnrc code system was used for all Monte Carlo calculations. BEAMnrc models of nine different linear accelerators with and without flattening filter were used to create realistic photon sources. Monte Carlo based calculations to determine the fluence perturbations due to the presens of the chambers components, the different material of the sensitive volume (air instead of water) as well as the volume effect were performed by the user code egs-chamber. Results: Stem, central electrode, wall, density and volume perturbation factors for linear accelerators with and without flattening filter were calculated as a function of the beam quality specifier TPR{sub 20/10}. A bias between the perturbation factors as a function of TPR{sub 20/10} for flattening filter free beams and conventional linear accelerators could not be observed for the perturbations caused by the components of the ionization chamber and the sensitive volume. Conclusion: The results indicate that the well-known small bias between the beam quality correction factor as a function of TPR20/10 for the flattening filter free and conventional linear accelerators is not caused by the geometry of the detector but rather by the material of the sensitive volume. This suggest that the bias for flattening filter free photon fields is only caused by the different material of the sensitive volume (air instead of water)

60 APPLIED LIFE SCIENCES↗

Characterization of cartridge filters from the IEA-R1 Nuclear Reactor

The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institute (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Transuranic Transmutation and Criticality Calculation Sensitivity to Heterogeneous Lattice Effects - 12391

Using Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies has been researched at length and has shown to provide the benefit of transmutation and targets the amount and toxicity of high level waste needed to be managed. Advanced MOX concepts using enriched Uranium Dioxide (UO{sub 2}) are required for multiple recycling of plutonium. The use of MOX and ordinary UO{sub 2} fuel in the same assembly as well as unfueled rods and assembly edge effects contrasts with the unit cell computational assumption of a uniform infinite array of rods. While a deterministic method of calculating the Dancoff factor has traditionally been employed in fuel assembly analysis due to the lighter computational and modeling requirements, this research seeks to determine the validity of the uniform, infinite lattice assumption with respect to Dancoff factor and determine the magnitude of the impact of nonuniform lattice effects on fuel assembly criticality calculations as well as transuranic isotope production and transmutation. This research explored the pin-to-pin interaction in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel rods through the impact of the calculated Dancoff factors from the deterministic method used in SCALE versus the Monte Carlo method used in the code DANCOFF-MC. Using the Monte Carlo method takes into account the non-uniform lattice effects of having neighboring fuel rods with different cross-sectional spectra whereas the Dancoff factor calculated by SCALE assumes a uniform, infinite lattice of one fuel rod type. Differences in eigenvalue calculations as a function of burnup are present between the two methods of Dancoff factor calculation. The percent difference is greatest at low burnup and then becomes smaller throughout the cycle. Differences in the transmutation rate of transuranic isotopes in the MOX fuel are also present between the Dancoff factor calculation methods. The largest difference is in Pu-239, Pu-242, and Am-241 composition whereas U-238, Pu-242, and Pu-238 composition was not changed by taking into account the non-homogenous lattice effects. Heterogeneous lattice effects do change the calculated eigenvalue and transmutation rate in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel. However, the uncertainty in the ENDF data used by SCALE in these calculations is large enough that the infinite lattice assumption remains valid. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

PIXE simulation: Models, methods and technologies

The simulation of PIXE (Particle Induced X-ray Emission) is discussed in the context of general-purpose Monte Carlo systems for particle transport. Dedicated PIXE codes are mainly concerned with the application of the technique to elemental analysis, but they lack the capability of dealing with complex experimental configurations. General-purpose Monte Carlo codes provide powerful tools to model the experimental environment in great detail, but so far they have provided limited functionality for PIXE simulation. This paper reviews recent developments that have endowed the Geant4 simulation toolkit with advanced capabilities for PIXE simulation, and related efforts for quantitative validation of cross sections and other physical parameters relevant to PIXE simulation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗