Results 71 to 80 of about 898 (171)

Opencg: A Combinatorial Geometry Modeling Tool for Data Processing and Code Verification [PDF]

, 2015
Combinatorial Geometry (CG) is one formulation for computational geometric models that is commonly used in many neutron transport simulation codes. The use of CG is advantageous since it permits an accurate yet concise representation of complex reactor ...
Boyd III, William Robert Dawson, Forget, Benoit Robert Yves, Smith, Kord S.   +2 more
core  

Benchmarking of the SPERT-III E-core experiment with the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC [PDF]

EPJ Web of Conferences
This paper presents a code-to-code verification of the SPERT-III reactor in its E-core configuration using the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC.
Dervaux Antoine, Kinast Shai, Lemaire Matthieu, Di Blasi Vincenzo, Larmier Coline, Mancusi Davide, Zoia Andrea   +6 more
doaj   +1 more source

Investigation of improved methods for assessing convergence of models in MCNP using Shannon entropy [PDF]

, 2012
Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012."June 2012." Cataloged from PDF version of thesis.Includes bibliographical references (p.
Macdonald, Ruaridh (Ruaridh R.)
core  

Using Fractional Cascading to Accelerate Cross Section Lookups in Monte Carlo Neutron Transport Calculations [PDF]

, 2015
We describe and test a technique for carrying out energy grid searches in continuous-energy Monte Carlo (MC) neutron transport calculations that represents an optimal compromise between grid search performance and memory footprint.
Forget, Benoit Robert Yves, Josey, Colin, Lund, Amanda L., Romano, Paul Kollath, Siegel, Andrew M.   +4 more
core  

Spatially Continuous Depletion Algorithm for Monte Carlo Simulations [PDF]

, 2016
To correctly predict reactor behavior during cycle operations, the evolution of nuclide number densities throughout the core must be accurately modeled.
Ellis, Matthew Shawn, Josey, Colin, Smith, Kord S.   +2 more
core  

Using OpenMC in simulations of a low dimensional cold neutron moderator for the ICONE project [PDF]

EPJ Web of Conferences
In this contribution we showcase the use of OpenMC for the validation and optimization of a low dimensional cold neutron moderator for the ICONE project. An overview of our simulation strategy is given and initial results for an optimized layout of a low-
Wagner Richard, Ott Frédéric, Darpentigny Jacques   +2 more
doaj   +1 more source

Exploring the Performance Benefit of Hybrid Memory System on HPC Environments

, 2017
Hardware accelerators have become a de-facto standard to achieve high performance on current supercomputers and there are indications that this trend will increase in the future.
Cicotti, P., Gioiosa, R., Kestor, G., Laure, Erwin,, Markidis, Stefano,, Peng, Ivy Bo,   +5 more
core   +1 more source

Verification of the 3D Method of characteristics solver in OpenMOC [PDF]

, 2016
The Method of Characteristics (MOC) has seen wide interest in full-core reactor physics analysis due to its computational efficiency and ability to easily treat complex geometries.
Forget, Benoit Robert Yves, Gunow, Geoffrey Alexander, Shaner, Samuel Christopher, Smith, Kord S.   +3 more
core  

Multigroup diffusion preconditioners for multiplying fixed-source transport problems [PDF]

, 2014
Several preconditioners based on multigroup di usion are developed for application to multiplying fi xed-source transport problems using the discrete ordinates method.
Forget, Benoit, Roberts, Jeremy A.
core   +1 more source

Methodology and preliminary verification of generating heterogeneous multigroup microscopic cross-section libraries for neutron transport codes based on OpenMC

Nuclear Engineering and Technology
Advancements in reactor technology, particularly Generation IV and modular reactors, have introduced new challenges on the neutronics analysis due to their complex geometries and spectra.
Bowen Cui, Guohua Chen, Xiaofeng Jiang
doaj   +1 more source