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Title: Visualization of geometry and tally data using MCNP and Justine

Abstract

The Monte Carlo N-Particle (MCNP) transport code is a general-purpose code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport, including the capability to calculate eigenvalues for neutron-multiplying systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Justine is the graphical user interface and problem setup tool for the Los Alamos Radiation Modeling Interactive Environment (LARAMIE). Its purpose is to serve as a convenient and very general interface for setting up physics calculations and linking together the disparate radiation transport codes under a single front-end. Currently, the LARAMIE system includes MCNP and the deterministic transport code suit DANTSYS (ONEDANT, TWODANT, and THREEDANT, for one-, two-, and three-dimensional geometries, respectively). Justine is currently available through the Radiation Safety Information Computational Center to members of the criticality safety community for evaluation and use. The authors will demonstrate the capabilities of both codes for visualization of geometries and results from a variety of criticality problems.

Authors:
;  [1]
  1. Los Alamos National Lab., NM (United States)
Publication Date:
OSTI Identifier:
678146
Report Number(s):
CONF-990605-
Journal ID: TANSAO; ISSN 0003-018X; TRN: 99:009131
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 80; Conference: 1999 annual meeting of the American Nuclear Society (ANS), Boston, MA (United States), 6-10 Jun 1999; Other Information: PBD: 1999
Country of Publication:
United States
Language:
English
Subject:
99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; 42 ENGINEERING NOT INCLUDED IN OTHER CATEGORIES; M CODES; RADIATION TRANSPORT; MONTE CARLO METHOD; COMPUTER GRAPHICS; MAN-MACHINE SYSTEMS; INTERACTIVE DISPLAY DEVICES; CRITICALITY

Citation Formats

Cox, L.J., and Favorite, J.A. Visualization of geometry and tally data using MCNP and Justine. United States: N. p., 1999. Web.
Cox, L.J., & Favorite, J.A. Visualization of geometry and tally data using MCNP and Justine. United States.
Cox, L.J., and Favorite, J.A. Wed . "Visualization of geometry and tally data using MCNP and Justine". United States.
@article{osti_678146,
title = {Visualization of geometry and tally data using MCNP and Justine},
author = {Cox, L.J. and Favorite, J.A.},
abstractNote = {The Monte Carlo N-Particle (MCNP) transport code is a general-purpose code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport, including the capability to calculate eigenvalues for neutron-multiplying systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Justine is the graphical user interface and problem setup tool for the Los Alamos Radiation Modeling Interactive Environment (LARAMIE). Its purpose is to serve as a convenient and very general interface for setting up physics calculations and linking together the disparate radiation transport codes under a single front-end. Currently, the LARAMIE system includes MCNP and the deterministic transport code suit DANTSYS (ONEDANT, TWODANT, and THREEDANT, for one-, two-, and three-dimensional geometries, respectively). Justine is currently available through the Radiation Safety Information Computational Center to members of the criticality safety community for evaluation and use. The authors will demonstrate the capabilities of both codes for visualization of geometries and results from a variety of criticality problems.},
doi = {},
journal = {Transactions of the American Nuclear Society},
number = ,
volume = 80,
place = {United States},
year = {1999},
month = {9}
}