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Title: MCNP/X Transport in the Tabular Regime

Abstract

We review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, we emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. We also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

Authors:
 [1]
  1. Los Alamos National Laboratory, Group X-3-MCC, MS A143, Los Alamos, NM 87545 (United States)
Publication Date:
OSTI Identifier:
21055007
Resource Type:
Journal Article
Resource Relation:
Journal Name: AIP Conference Proceedings; Journal Volume: 896; Journal Issue: 1; Conference: Hadronic shower simulation workshop, Batavia, IL (United States), 6-8 Sep 2006; Other Information: DOI: 10.1063/1.2720460; (c) 2007 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSICS; 46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; COMPUTERIZED SIMULATION; CROSS SECTIONS; ELECTRONS; M CODES; MONTE CARLO METHOD; NEUTRONS; PHOTON TRANSPORT; PHOTONS; PROTON TRANSPORT; REVIEWS; TRANSPORT THEORY

Citation Formats

Hughes, H. Grady. MCNP/X Transport in the Tabular Regime. United States: N. p., 2007. Web. doi:10.1063/1.2720460.
Hughes, H. Grady. MCNP/X Transport in the Tabular Regime. United States. doi:10.1063/1.2720460.
Hughes, H. Grady. Mon . "MCNP/X Transport in the Tabular Regime". United States. doi:10.1063/1.2720460.
@article{osti_21055007,
title = {MCNP/X Transport in the Tabular Regime},
author = {Hughes, H. Grady},
abstractNote = {We review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, we emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. We also briefly touch on the current situation in regard to photon, electron, and proton transport tables.},
doi = {10.1063/1.2720460},
journal = {AIP Conference Proceedings},
number = 1,
volume = 896,
place = {United States},
year = {Mon Mar 19 00:00:00 EDT 2007},
month = {Mon Mar 19 00:00:00 EDT 2007}
}
  • The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.
  • The US Department of Energy Fissile Materials Disposition Program has begun studies for disposal of surplus weapons-grade plutonium (WG-Pu) as mixed uranium-plutonium oxide (MOX) fuel for commercial light water reactors (LWRs). Most MOX fuel experience is with reactor-grade plutonium (RG-Pu). Therefore, to use WG-Pu in MOX fuel, one must demonstrate that the experience with RG-Pu is relevant. Initial tests have been made in an I-hole of the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL) to aid in the investigation of some of the unresolved issues. One of these issues is to understand the impactmore » of gallium on LWR MOX fuel performance since it is present in small amounts in WG-Pu. Initial radiation transport calculations of the test specimens have been made at INEEL using the MCNP Monte Carlo radiation transport code. These calculations were made to determine the linear heating rates in the fuel specimens. Because of the nature of Monte Carlo, it is extremely time consuming and inefficient to show detailed hot spots in the specimens. However, results from discrete ordinates radiation transport calculations could show these spatial details. Therefore, INEEL was tasked with producing an MCNP source at the boundary of a rectangular parallel-piped enclosing the ATR I-hole, and Oak Ridge National Laboratory (ORNL) was tasked with transforming this boundary source into a discrete ordinates boundary source for the Three dimensional Oak Ridge radiation Transport (TORT) code. The results of this work are discussed.« less
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