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Title: Certification of MCNP Version 4A for WHC computer platforms. Revision 7

Abstract

MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).

Authors:
Publication Date:
Research Org.:
Westinghouse Hanford Co., Richland, WA (United States)
Sponsoring Org.:
USDOE, Washington, DC (United States)
OSTI Identifier:
71572
Report Number(s):
WHC-SD-MP-SWD-30001-Rev.7
ON: DE95011546; TRN: 95:015321
DOE Contract Number:
AC06-87RL10930
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 3 May 1995
Country of Publication:
United States
Language:
English
Subject:
66 PHYSICS; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; NEUTRAL-PARTICLE TRANSPORT; M CODES; MONTE CARLO METHOD; NEUTRON TRANSPORT; PHOTON TRANSPORT; EIGENVALUES; THREE-DIMENSIONAL CALCULATIONS; MANUALS; COMPUTER ARCHITECTURE; TORI

Citation Formats

Carter, L.L. Certification of MCNP Version 4A for WHC computer platforms. Revision 7. United States: N. p., 1995. Web. doi:10.2172/71572.
Carter, L.L. Certification of MCNP Version 4A for WHC computer platforms. Revision 7. United States. doi:10.2172/71572.
Carter, L.L. Wed . "Certification of MCNP Version 4A for WHC computer platforms. Revision 7". United States. doi:10.2172/71572. https://www.osti.gov/servlets/purl/71572.
@article{osti_71572,
title = {Certification of MCNP Version 4A for WHC computer platforms. Revision 7},
author = {Carter, L.L.},
abstractNote = {MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).},
doi = {10.2172/71572},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Wed May 03 00:00:00 EDT 1995},
month = {Wed May 03 00:00:00 EDT 1995}
}

Technical Report:

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  • MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).
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