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At least 343 records · Page 19

Simulation of Infrared Laser Heating of Silica Using Heat Conduction and Multifrequency Radiation Diffusion Equations Adapted for Homogeneous Refractive Lossy Media

Localized, transient heating of materials using micro-scale, highly absorbing laser light has been used in many industries to anneal, melt and ablate material with high precision. Accurate modeling of the relative contributions of conductive, convective and radiative losses as a function of laser parameters is essential to optimizing micro-scale laser processing of materials. In bulk semi-transparent materials such as silicate glass melts, radiation transport is known to play a significantly larger role as the temperature increases. Conventionally, radiation is treated in the frequency-averaged diffusive limit (Rosseland approximation). However, the role and proper treatment of radiative processes under rapidly heated, high thermal gradient conditions, often created through laser-matter interactions, is at present not clear. Starting from the radiation transport equation for homogeneous, refractive lossy media, they derive the corresponding time-dependent multi-frequency diffusion equations. Zeroth and first moments of the transport equation couple the energy density, flux and pressure tensor. The system is closed by neglecting the temporal derivative of the flux and replacing the pressure tensor by its diagonal analogue. The radiation equations are coupled to a diffusion equation for the matter temperature. They are interested in modeling infrared laser heating of silica over sub-millimeter length scales, and at possibly rapid rates. Hence, in contrast to related work, they retain the temporal derivative of the radiation field. They derive boundary conditions at a planar air-silica interface taking account of reflectivities obtained from the Fresnel relations that include absorption. The effect of a temperature-dependent absorption index is explored through construction of a multi-phonon dielectric function that includes mode dispersion. The spectral dimension is discretized into a finite number of intervals yielding a system of multigroup diffusion equations. Simulations are presented. To demonstrate the bulk heat loss due to radiation and the effect of the radiation's temporal derivative, they model cooling of a silica slab, initially at 2500 K, for 10 s. Retaining the derivative enables correctly modeling the loss of photons initially present in the slab. Other simulations model irradiating silica discs (of approximately 5 mm radii and thickness) with a CO2 laser: {lambda} = 10.59 and 4.6 um, Gaussian profile, r{sub 0} = 0.5 mm for 1/e decay. By surrounding the disks in room-temperature air, they make use of the boundary conditions described above.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Monte Carlo Criticality Methods and Analysis Capabilities in SCALE

This paper describes the Monte Carlo codes KENO V.a and KENO-VI in SCALE that are primarily used to calculate multiplication factors and flux distributions of fissile systems. Both codes allow explicit geometric representation of the target systems and are used internationally for safety analyses involving fissile materials. KENO V.a has limiting geometric rules such as no intersections and no rotations. These limitations make KENO V.a execute very efficiently and run very fast. On the other hand, KENO-VI allows very complex geometric modeling. Both KENO codes can utilize either continuous-energy or multigroup cross-section data and have been thoroughly verified and validated with ENDF libraries through ENDF/B-VII.0, which has been first distributed with SCALE 6. Development of the Monte Carlo solution technique and solution methodology as applied in both KENO codes is explained in this paper. Available options and proper application of the options and techniques are also discussed. Finally, performance of the codes is demonstrated using published benchmark problems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Application of the INSTANT-HPS PN Transport Code to the C5G7 Benchmark Problem

INSTANT is the INL's next generation neutron transport solver to support high-fidelity multi-physics reactor simulation INSTANT is in continuous development to extend its capability Code is designed to take full advantage of middle to large cluster (10-1000 processors) Code is designed to focus on method adaptation while also mesh adaptation will be possible. It utilizes the most modern computing techniques to generate a neutronics tool of full-core transport calculations for reactor analysis and design. It can perform calculations on unstructured 2D/3D triangular, hexagonal and Cartesian geometries. Calculations can be easily extended to more geometries because of the independent mesh framework coded with the model Fortran. This code has a multigroup solver with thermal rebalance and Chebyshev acceleration. It employs second-order PN and Hybrid Finite Element method (PNHFEM) discretization scheme. Three different in-group solvers - preconditioned Conjugate Gradient (CG) method, preconditioned Generalized Minimal Residual Method (GMRES) and Red-Black iteration - have been implemented and parallelized with the spatial domain decomposition in the code. The input is managed with extensible markup language (XML) format. 3D variables including the flux distributions are outputted into VTK files, which can be visualized by tools such as VisIt and ParaView. An extension of the code named INSTANTHPS provides the capability to perform 3D heterogeneous transport calculations within fuel pins. C5G7 is an OECD/NEA benchmark problem created to test the ability of modern deterministic transport methods and codes to treat reactor core problems without spatial homogenization. This benchmark problem had been widely analyzed with various code packages. In this transaction, results of the applying the INSTANT-HPS code to the C5G7 problem are summarized.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Resonance Self-Shielding Methodologies in SCALE 6

SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

VERIFICATION OF THE SHIFT MONTE CARLO CODE

Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a fully-functional parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift s Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP

42 ENGINEERING↗

Enhancements in SCALE 6.1

The SCALE code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a 'plug-and-play' framework with 89 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.1 builds on the existing capabilities and ease-of-use of SCALE and provides several new features such as enhanced lattice physics capabilities and multigroup Monte Carlo depletion, improved options and capabilities for sensitivity and uncertainty analysis calculations, improved flexibility in shielding and criticality accident alarm system calculations with automated variance reduction, and new options for the definition of group structures for depletion calculations. The SCALE 6.1 development team has focused on improved robustness via substantial additional regression testing and verification for new and existing features, providing improved performance relative to SCALE 6.0, especially in reactor physics calculations and in the nuclear data used for source term characterization and shielding calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The Embedded Self-Shielding Method

The embedded self-shielding method (ESSM) is described for computing resonance-shielded cross sections used in multigroup neutron transport calculations with the SCALE code system. The ESSM - embeds the self-shielding computation within the transport solution. The transport solution provides information for treating heterogeneous self-shielding effects, and the resulting shielded cross sections are fed back to the transport calculation. Iterations are done to obtain self-consistency. This allows self-shielded cross sections to be generated directly in the transport geometry without requiring external computation of Dancoff factors. The ESSM theory and example calculations are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of a Statistical Sampling Method for Uncertainty Analysis with SCALE

A new statistical sampling sequence called Sampler has been developed for the SCALE code system. Random values for the input multigroup cross sections are determined by using the XSUSA program to sample uncertainty data provided in the SCALE covariance library. Using these samples, Sampler computes perturbed self-shielded cross sections and propagates the perturbed nuclear data through any specified SCALE analysis sequence, including those for criticality safety, lattice physics with depletion, and shielding calculations. Statistical analysis of the output distributions provides uncertainties and correlations in the desired responses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Use of Covariance Matrices in a Consistent (Multiscale) Data Assimilation for Improvement of Basic Nuclear Parameters in Nuclear Reactor Applications: from Meters to Femtometers

A new approach is proposed, the consistent data assimilation, that allows to link the integral data experiment results to basic nuclear parameters employed by evaluators to generate ENDF/B point energy files in order to improve them. A practical example is provided where the sodium neutron propagation experiments, EURACOS and JANUS-8, are used to improve via modifications of {sup 23}Na nuclear parameters (like scattering radius, resonance parameters, Optical model parameters, Statistical Hauser-Feshbach model parameters, and Preequilibrium Exciton model parameters) the agreement of calculation versus experiments for a series of measured reaction rate detectors slopes. Future work involves comparison of results against a more traditional multigroup adjustments, and extension to other isotope of interest in the reactor community as {sup 56}Fe, actinides, and fission products.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Rayleigh Quotient Iteration in 3D, Deterministic Neutron Transport

Today's "grand challenge" neutron transport problems require 3-D meshes with billions of cells, hundreds of energy groups, and accurate quadratures and scattering expansions. Leadership-class computers provide platforms on which high-fidelity fluxes can be calculated. However, appropriate methods are needed that can use these machines effectively. Such methods must be able to use hundreds of thousands of cores and have good convergence properties. Rayleigh quotient iteration (RQI) is an eigenvalue solver that has been added to the Sn code Denovo to address convergence. Rayleigh quotient iteration is an optimal shifted inverse iteration method that should converge in fewer iterations than the more common power method and other shifted inverse iteration methods for many problems of interest. Denovo's RQI uses a new multigroup Krylov solver for the fixed source solutions inside every iteration that allows parallelization in energy in addition to space and angle. This Krylov solver has been shown to scale successfully to 200,000 cores: for example one test problem scaled from 69,120 cores to 190,080 cores with 98% efficiency. This paper shows that RQI works for some small problems. However, the Krylov method upon which it relies does not always converge because RQI creates ill-conditioned systems. This result leads to the conclusion that preconditioning is needed to allow this method to be applicable to a wider variety of problems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON

The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A Spectral Verification of the HELIOS-2 Lattice Physics Code

Core modeling of the Advanced Test Reactor (ATR) at INL is currently undergoing a significant update through the Core Modeling Update Project1. The intent of the project is to bring ATR core modeling in line with today’s standard of computational efficiency and verification and validation practices. The HELIOS-2 lattice physics code2 is the lead code of several reactor physics codes to be dedicated to modernize ATR core analysis. This presentation is concerned with an independent verification of the HELIOS-2 spectral representation including the slowing down and thermalization algorithm and its data dependency. Here, we will describe and demonstrate a recently developed simple cross section generation algorithm based entirely on analytical multigroup parameters for both the slowing down and thermal spectrum. The new capability features fine group detail to assess the flux and multiplication factor dependencies on cross section data sets using the fundamental infinite medium as an example.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Monte Carlo Calculations of Eigenvalue Sensitivities to System Dimensions

A Monte Carlo method is developed that computes sensitivities or derivatives of the k-eigenvalue with respect to interface boundary locations for uniform expansions or contractions. The method is implemented in a research version of MCNP6, and its results are compared with reference solutions obtained from equivalent multigroup calculations with PARTISN (discrete ordinates) and direct Monte Carlo perturbations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of high-temperature gas-cooled reactor configurations. This paper describes part of the results of that effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data and the MCNP predictions.

code validation↗

Overview of Scale 6.2

SCALE is an industry-leading suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a plug-and-play framework that includes three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 provides several new capabilities and significant improvements in many existing features, especially with expanded CE Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. A brief overview of SCALE capabilities is provided with emphasis on new features for SCALE 6.2.

97 MATHEMATICS AND COMPUTING↗

Use of SCALE Continuous-Energy Monte Carlo Tools for Eigenvalue Sensitivity Coefficient Calculations

The TSUNAMI code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The CLUTCH and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE KENO framework to generate the capability for TSUNAMI-3D to perform eigenvalue sensitivity calculations in continuous-energy applications. This work explores the improvements in accuracy that can be gained in eigenvalue and eigenvalue sensitivity calculations through the use of the SCALE CE KENO and CE TSUNAMI continuous-energy Monte Carlo tools as compared to multigroup tools. The CE KENO and CE TSUNAMI tools were used to analyze two difficult models of critical benchmarks, and produced eigenvalue and eigenvalue sensitivity coefficient results that showed a marked improvement in accuracy. The CLUTCH sensitivity method in particular excelled in terms of efficiency and computational memory requirements.

97 MATHEMATICS AND COMPUTING↗

Enhancements in Continuous-Energy Monte Carlo Capabilities for SCALE 6.2

SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, industry, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a plug-and-play framework that includes three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 provides several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, sensitivity and uncertainty analysis, and improved fidelity in nuclear data libraries. A brief overview of SCALE capabilities is provided with emphasis on new features for SCALE 6.2.

Continuous-Energy↗

Sandia Computational Engine for Particle Transport for Radiation Effects v 1.4

The SCEPTRE code solves the linear Boltzmann transport equation for general two- and three-dimensional geometries. SCEPTRE is capable of handling any particle type for which multigroup-Legendre cross sections are available. However, the code is designed primarily to model the transport of photons, electrons, and positrons through matter.

Source record↗