Graphical User Input Interface for MCNP
- Westinghouse Hanford Co., Richland, WA (United States)
The Monte Carlo Neutron Photon (MCNP) computer code is a transport code that includes a powerful three-dimensional geometry and source modeling capability. MCNP has applications in shielding, reactor physics, criticality, and many other areas. In creating an input file, entering laborious descriptions of geometry, sources, tallies, and optimization parameters with a line editor is very tedious and error prone. Input files for MCNP typically contain from hundreds to thousands of lines of data. This makes it difficult during the line-edit sessions to remember exactly what portion of the input file still remains to be completed. Once an input file is created, additional time is required to plot the geometry and correct any errors in the geometry description. To aid in the creation of MCNP input files, a graphical interface has been developed that enables the use to interactively build the geometry with the aid of two or more dynamic cross-sectional views of the model.
- Research Organization:
- Westinghouse Hanford Co., Richland, WA (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- OSTI ID:
- 6844637
- Report Number(s):
- CONF-931160--
- Journal Information:
- Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 69; ISSN TANSAO; ISSN 0003-018X
- Publisher:
- American Nuclear Society
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
220100* -- Nuclear Reactor Technology-- Theory & Calculation
97 MATHEMATICS AND COMPUTING
990200 -- Mathematics & Computers
CALCULATION METHODS
COMPUTER CODES
COMPUTER GRAPHICS
CRITICALITY
DATA PROCESSING
Graphical Interface
M CODES
MONTE CARLO METHOD
Monte Carlo Neutron Photon (MCNP) Computer Code
Nuclear Criticality Safety Program (NCSP)
PHYSICS
PROCESSING
REACTOR PHYSICS
SHIELDING
THREE-DIMENSIONAL CALCULATIONS