Visualization of geometry and tally data using MCNP and Justine
- Los Alamos National Lab., NM (United States)
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.
- OSTI ID:
- 678146
- Report Number(s):
- CONF-990605-; ISSN 0003-018X; TRN: 99:009131
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 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
Similar Records
Justine user`s manual
A primer for criticality calculations with DANTSYS