SEARCH · Search NASA

Results for “Constructive solid geometry”

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.

SourceResearch connectionFrom yearThrough year

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

At least 37 records · Page 2

All sources NASA NTRS DOE OSTI

DOE OSTI2018

The MCNP6 Book On Unstructured Mesh Geometry: User's Guide for MCNP 6.2.1

Los Alamos National Laboratory’s (LANL) Monte Carlo N-Particle R (MCNP R ) transport code has a more general geometry capability than has been available in most combinatorial geometry codes [1]. In addition to the capability of combining several predefined geometric bodies, as in a combinatorial scheme, MCNP6 [2] gives the user the added flexibility of defining geometric regions from all the first and second degree surfaces of analytical geometry and elliptical tori and then combining them with Boolean operators. This decades-old constructive solid geometry (CSG) capability has been well-tested and verified. However, it has long been recognized that as the model complexity increases, this process of describing the geometry is difficult, tedious, time-consuming, and error prone[3, 4, 5]. Consequently, innovators have taken on the task of developing a better way to construct geometries, not only in MCNP6, but other particle transport codes as well

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

Development of Fusion Neutronics Methodology, Software, and Testing Facility in FDS Team

Fusion neutronics including methodology, software, experiments, etc. is studied concerning the generation, utilizing and protection of neutrons in fusion facilities. Achievements have been made in the past decades to solve the new problems fusion neutronics. However there is still some way to go such as following. Automatic modeling method based on CAD has been developed to avoid the heavy burden of manual geometry description. The pre-processing costs much time to simplify the CAD model for converting to Constructive Solid Geometry (CSG), 2D or 3D mesh based geometry. This problem has become a new focus. Deep penetration of radiation shielding is notable problem for fusion. Adaptive variance reduction methods such as Consistent Adjoint Driven Importance Sampling (CADIS) were developed to accelerate convergence. However the gap still exists between engineering requirements and current methodology. The geometry simplification, homogenizing based tally, error in nuclear data and error propagation in coupled calculation, etc., causes considerable uncertainty, which leads to 'safety factor' applied in simulation results to achieve required safety margin. It can significantly raise the cost and even makes it impractical to initiate the fusion facilities. Uncertainty qualification, error analysis and error propagation problem should be studied. New methodology and software should be validated. For example the (n,p), (n,a), (n,t) cross section of Cu, Fe, W, Ni which is important for evaluation of material radiation damage is absent. The validation of neutronics properties of special components such as first wall, blanket and coil are needed. The D-T fusion neutron generator can directly represent the neutron environment in fusion reactors. However the neutron yield is several orders of magnitudes lower than requirements. To resolve obstacles in fusion neutronics, broad improvements of methodology, software and testing facility have been achieved by FDS Team in recent years and the main progress was introduced in this paper.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Critical Experiment Benchmark Results using MCNP6.2 Unstructured Mesh

The purpose of this report is to document the results of five critical experiment benchmarks modeled using the unstructured mesh (UM) geometry capability in MCNP®1 code, version 6.2. LANL currently has a vast critical experiment benchmark library with over 1100 cases, modeled using constructive solid geometry (CSG) and packaged with Whisper-1.1. Validation efforts for the purposes of nuclear criticality safety typically focus on benchmark libraries built using CSG models. Results of UM geometry models are compared with CSG results and the critical experiment k-effective from the ICSBEP handbook. All computed k-effective results in this study are well within 1% of the experimental result.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Layered CAD/CSG geometry for spatially complex radiation transport scenarios

Many spatially complex fission, fusion, and national security Monte Carlo (MC) radiation transport scenarios involve combining computer-aided design (CAD) models with constructive solid geometry (CSG) models. A layered geometry method has been implemented in the Shift MC code to address this need. With layered geometry, multiple CAD and/or CSG models can be clipped, translated, rotated, and placed in overlapping layers to form transport-ready geometries. Here, the utility of this method is demonstrated with two problems: (1) a fixed-source simulation with a layered geometry consisting of a LiDAR-generated CAD model of the Combined Arms Collective Training Facility urban environment overlaid with CSG models of a mock hotel and a detector apparatus, and (2) a k-eigenvalue calculation using a layered geometry model of the Transformational Challenge Reactor consisting of CAD fuel elements placed in a CSG core. Tallied particle flux distributions match expectations, but tracking robustness must be improved prior to general-purpose use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

MCNP6.3 Unstructured Mesh Verification: GodivR and CANDU Models

A geometric cell of the Monte Carlo N-Particle (MCNP)1 transport code is traditionally created by using Boolean operators on defined surfaces. This constructive solid geometry (CSG) capability has been available in the MCNP code since its beginning. However, a CSG model approach is limited when it comes to constructing a representative geometry for a complex model in its ability to capture a correct model representation. Starting with the version 6.0, the MCNP code has the ability of embedding an unstructured mesh (UM) model into a CSG cell to create a hybrid geometry [1]. The MCNP UM feature provides the flexibility of defining very complex geometries because computer aided design (CAD) and mesh generation software packages can be utilized to construct UM models for MCNP simulations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

In Situ Visualization of Radiation Transport Geometry

The ultimate goal for radiation transport is to perform full-core reactor modelling and simulation. Advances in computational simulation bring this goal close to reality and the newest Monte Carlo transport codes have begun to shift to using accelerators that have become a stalwart in the supercomputing and HPC space. Within the reactor modelling and simulation community, Monte Carlo transport simulations are considered the gold standard for simulation. Through the use of "combinatorial geometry" (constructive solid geomtry), complex models can be used with fewer approximation compromises while at the same time scale to run on some of the largest supercomputers in the world.Unfortunately, the state-of-the-art for "combinatorial geometry" visualization is to decompose the geometry into a mesh. This approach could require a significant amount of memory which is antithetical to in situ visualization. To address this issue, we introduce a ray caster for visualizing combinatorial geometry in radiation transport code. By only using the accelerators for the radiation transport code and leaving the CPU cores idle, there is an opportunity to conduct on node in situ visualization with the idle CPU cores, something domain experts have up to this point been unable to do. By utilizing VTK-m, the visualization can be run on the CPU as this particular application demands, but also run on any architecture that is supported by VTK-m, enabling future re-use across different platforms.

Kim, Mark B.↗

DOE OSTI2023

An unstructured mesh based neutronics optimization workflow

We have developed a fully automated workflow to optimize the neutronics performance of the Second Target Station (STS) at the Oak Ridge National Laboratory’s Spallation Neutron Source. The optimization workflow starts with the parametrized solid CAD engineering models and converts them into the unstructured mesh (UM) models for the neutronics calculations with MCNP6.2. Calculations are executed and their results are loaded into the Dakota optimization toolkit. Dakota analyzes the results and proposes new geometry parameters for the next design iteration. The cycle repeats until the optimal parameters are found. The automated CAD to MCNP conversion, the use of high-fidelity UM models, and the use of modern optimizer are the key elements that advance the entire optimization workflow in comparison with the original workflow. The original workflow was based on a simplified constructive solid geometry (CSG) modeling with MCNPX, mcnp_pstudy tool, and an in-house optimizer. Herein to demonstrate the new workflow, we present a case of neutronics optimization of the moderator–reflector assembly (MRA). Apart from the MRA, the workflow can optimize other major STS components, such as the spallation target, neutron beamlines, radiation shielding, and various accelerator components. Importantly, the new workflow opens the door to the advanced multi-physics multi-parameter optimization and has the potential for use in other nuclear physics and accelerator applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Surface Triangulation for CSG in Mercury

Visualization routines for rendering complicated geometries are very useful for engineers and scientists who are trying to build 3D prototypes of their designs. A common way to rapidly add interesting features to a 3D model is through the use of a concept called Constructive Solid Geometry. CSG uses compositions of the boolean set operations to manipulate basic geometric primitives to form more complicated objects. The most common boolean operations employed are union, intersection, and subtraction. Most computer-aided design software packages contain some sort of ability visualize CSG. The typical workflow for the user is as follows: The user specifies the individual primitive components, the user arbitrarily combines each of these primitives with boolean operations, the software generates a CSG tree structure which normally stores these solids implicitly with their defining equation, the tree is traversed and a general algorithm is applied to render the appropriate geometry onto the screen. Algorithms for visualizing CSG have been extensively developed for over a decade. Points sampled from the implicit solids are typically used as input by variations of algorithms like marching cubes and point-cloud surface reconstruction. Here, we explain a surface triangulation method from the graphics community that is being used for surface visualization in the framework of a Monte-Carlo neutron transport code called Mercury.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2019

AFIT and LANL Collaboration Research Proposals [Slides]

MCNP is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP has the capability to track particles on unstructured polyhedral meshes that are embedded into the legacy constructive solid geometry (CSG).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Preliminary study on applying the method of characteristics to pebble-bed HTR - 18637

Neutron transport in high temperature pebble-bed gas-cooled reactor, which is a promising generation IV reactor candidate with cylindrical core, graphite moderator and spherical TRISO coated particles fuel, is of great interest. However, the stochastic method namely Monte Carlo faces with tremendous simulation of particles and the conventional two-step method featured by homogenization only can give the scalar flux. In this paper, method of characteristics is trying to be applied to pebble bed reactor because of its advantages in arbitrary geometry and the ability to obtain solution in neutronically large domains. Some challenges occur when it comes to HTR. First, the amount of calculation and the memory requirement are huge due to the large quantities of spherical fuel and stochastic distribution of pebbles. Second, the arrangement of characteristics and ray tracing process is more complex comparing to LWR with cuboid geometry because the stochastic distribution of the spherical assemblies and the longer neutron mean free path in pebble bed HTR. Third, the treatment of the multiscale sphere fuel need be detail studied to determine the subdivision of flat source region. At present, preliminary research on this subject is undergoing. This paper is concentrated on the modeling of the pebble-bed reactor core by constructive solid geometry (CSG) which allows complex spatial models to be built using boolean operations, characteristics arrangement and the treatment of boundary condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

FY25 MOOSE Usability Improvements: 3D Meshing Capabilities, Initiation of Geometry Support for Monte Carlo Tools, and Enhancement of MOOSE/Workbench User Input Interactions

Usability improvements have been made to MOOSE and Workbench in FY25 to enhance usability and user workflows. Assorted enhancement have been made to MOOSE’s intrinsic meshing capabilities in order to enable more flexible and complex meshing of nuclear reactor systems, in particular for 3D applications. Mesh generators have been added to perform operations such as batch mesh generation, surface mesh generation, and creation of 3D transition layers. These mesh generation capabilities make it much easier to generate high quality non-extruded 3D meshes. Additionally, work to integrate Monte Carlo reactor physics simulations into MOOSE-based multi-physics workflows has reached another milestone with the implementation of the Constructive Solid Geometry (CSG) base framework. This framework lays the foundation for mesh generators to offer the user a generic CSG output option (as opposed to a finite element mesh). To support users, workshop on the MOOSE Reactor Module was delivered which featured hands-on examples using the NEAMS Workbench on INL’s High Performance Computing system. Recent updates to the NEAMS Workbench, WASP, and the MOOSE language server have introduced several improvements aimed at making MOOSE-based simulation setup and input management faster, more accurate, and easier to use. Key capabilities that have been added include multi-tab-stop autocompletion, visual input diagnostics, developer-directed data visualizations, upgraded ParaView integration, and Workspace-level file tracking. Together, these changes make it easier for users to build, validate, and manage complex MOOSE-based simulation models — especially those involving reusable components, included files, and datasets. The improvements are designed to save time, reduce input errors, and help users get to a successful simulation run faster, with more confidence in the results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Kip, Version 1.0

This high-performance ray tracing library provides very fast rendering, compact code, type flexibility via C++ 'generic programming' techniques, and ease of use through an application programming interface (API) that functions independently of any GUI, on-screen display, or other enclosing application. Kip supports constructive solid geometry models based on any of several built-in shapes and operators and allows for new, user-defined shapes and operators. Additional features include basic texturing, input/output of models using a simple human-readable file format with full error checking and detailed diagnostics, and support for shared data parallelism. Kip is written in pure, ANSI Standard C++, is entirely platform independent, and is very easy to use, as it requires no build system, configuration or installation scripts, wizards, special preprocessing, makefiles, or external libraries.

Staley, Martin↗

DOE OSTI2018

Implementation and Validation of Photon Transport in OpenMC

OpenMC is a Monte Carlo particle transport code focused on reactor physics calculations. It stochastically simulates neutrons moving through 3D models using constructive solid geometry and continuous-energy cross sections. Recently, a photon transport capability was added to OpenMC. In reactor simulations, photon transport is needed, for example, to account for photon heating: although the majority of the energy deposited at a fission site comes from the kinetic energy of the fission fragments, a nontrivial portion can be carried away by photons. Therefore, photons produced in neutron reactions must be simulated in order to accurately model energy deposition. OpenMC can now simulate neutron-induced photon production as well as perform pure photon calculations. The four basic photon interactions with matter — coherent (Rayleigh) scattering, incoherent (Compton) scattering, the photoelectric effect, and pair production — are modeled. Secondary processes that can create new photons (i.e., atomic relaxation, electron-positron annihilation, and bremsstrahlung) are also included. The methods implemented here were validated against MCNP6 for a variety of materials and energies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Two MCNP Models for Computational Performance Benchmarking

The purpose of this report is to describe two non-trivial MCNP models and execution configurations that are suitable to characterizing computational performance. To that end, this report uses a constructive solid geometry (CSG) representation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) [1–3] to perform a k-eigenvalue calculation and an unstructured mesh (UM) representation of the International Commission on Radiological Protection (ICRP) publication 145 (ICRP145) male human phantom [4, 5] to perform a fixed-source calculation assuming that a 1 MeV photon source is distributed throughout the phantom’s liver.

97 MATHEMATICS AND COMPUTING↗

NASA NTRS1985

Improving the Effectiveness of Integral Property Calculation in a CSG Solid Modeling System by Exploiting Predictability

Integral property calculation is an important application for solid modeling systems. Algorithms for computing integral properties for various solid representation schemes are fairly well known. It is important to deigners and users of solid modeling systems to understand the behavior of such algorithms. Specifically the trade-off between execution time and accuracy is critical to effective use of integral property calculation. The average behavior of two algorithms for Constructive Solid Geometry (CSG) representations is investigated. Experimental results from the PADL-2 solid modeling system show that coarse decompositions can be used to predict execution time and error estimates for finer decompositions. Exploiting this predictability allow effective use of the algorithms in a solid modeling system.

Clark, A. L.↗

NASA NTRS1987

Octree based automatic meshing from CSG models

Finite element meshes derived automatically from solid models through recursive spatial subdivision schemes (octrees) can be made to inherit the hierarchical structure and the spatial addressability intrinsic to the underlying grid. These two properties, together with the geometric regularity that can also be built into the mesh, make octree based meshes ideally suited for efficient analysis and self-adaptive remeshing and reanalysis. The element decomposition of the octal cells that intersect the boundary of the domain is emphasized. The problem, central to octree based meshing, is solved by combining template mapping and element extraction into a procedure that utilizes both constructive solid geometry and boundary respresentation techniques. Boundary cells that are not intersected by the edge of the domain boundary are easily mapped to predefined element topology. Cells containing edges (and vertices) are first transformed into a planar polyhedron and then triangulated via element extractors. The modeling environments required for the derivation of planar polyhedra and for element extraction are analyzed.

Perucchio, Renato↗

NASA NTRS1987

Geometrical and topological issues in octree based automatic meshing

Finite element meshes derived automatically from solid models through recursive spatial subdivision schemes (octrees) can be made to inherit the hierarchical structure and the spatial addressability intrinsic to the underlying grid. These two properties, together with the geometric regularity that can also be built into the mesh, make octree based meshes ideally suited for efficient analysis and self-adaptive remeshing and reanalysis. The element decomposition of the octal cells that intersect the boundary of the domain is discussed. The problem, central to octree based meshing, is solved by combining template mapping and element extraction into a procedure that utilizes both constructive solid geometry and boundary representation techniques. Boundary cells that are not intersected by the edge of the domain boundary are easily mapped to predefined element topology. Cells containing edges (and vertices) are first transformed into a planar polyhedron and then triangulated via element extractor. The modeling environments required for the derivation of planar polyhedra and for element extraction are analyzed.

Saxena, Mukul↗

DOE OSTI2017

Investigations into alternative radiation transport codes for ITER neutronics analysis

As radiation transport models for ITER become increasingly complex, the use of traditional constructive solid geometry (CSG) modelling approaches such as MCWNP presents significant disadvantages during model creation and integration. Studies were performed to assess the capabilities of alternative mesh-based neutronics workflows offered by the MCNP, DAG-MCNP and Serpent-2 codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗