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Title: MCNP/X TRANSPORT IN THE TABULAR REGIME

Abstract

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.

Authors:
 [1]
  1. Los Alamos National Laboratory
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
OSTI Identifier:
1000494
Report Number(s):
LA-UR-07-0080
TRN: US1100136
DOE Contract Number:
AC52-06NA25396
Resource Type:
Conference
Resource Relation:
Conference: HADRONIC SHOWER SIMULATION WORKSHOP (HSSO6) ; 200609 ; BATAVIA
Country of Publication:
United States
Language:
English
Subject:
73; AVAILABILITY; CROSS SECTIONS; NEUTRONS; PROTON TRANSPORT; SIMULATION; TRANSPORT

Citation Formats

HUGHES, H. GRADY. MCNP/X TRANSPORT IN THE TABULAR REGIME. United States: N. p., 2007. Web.
HUGHES, H. GRADY. MCNP/X TRANSPORT IN THE TABULAR REGIME. United States.
HUGHES, H. GRADY. Mon . "MCNP/X TRANSPORT IN THE TABULAR REGIME". United States. doi:. https://www.osti.gov/servlets/purl/1000494.
@article{osti_1000494,
title = {MCNP/X TRANSPORT IN THE TABULAR REGIME},
author = {HUGHES, H. GRADY},
abstractNote = {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.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 08 00:00:00 EST 2007},
month = {Mon Jan 08 00:00:00 EST 2007}
}

Conference:
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  • 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.
  • Fluorescence yields for a series of elements using the gas-flow proportional detector, scintillation detector, and the universal detector are tabulated. (P.C.H.)
  • Monte Carlo methods have long been used at the Savannah River Laboratory (SRL) to perform criticality calculations for many different processes. To perform transport analyses (both neutron and photon), the two-dimensional infinite lattice integral transport code GLASS has been used. The neutron transport portion of the code has been benchmarked against other codes and experimental data. The photon transport portion of the code, which is used to calculate gamma redistribution in the event of a loss of moderator and/or coolant, had not been benchmarked against either. For this reason, the Monte Carlo code MCNP was used to benchmark the photonmore » transport portion of the GLASS code. Preceding this, a brief description of the geometry of the Savannah River Plant's (SRP's) reactor cores and how they were modeled using MCNP is given.« less
  • Previous work on code benchmarks has focused on numerical comparisons to experiments or other codes. Recently, analytical benchmarks have been applied to codes that solve the Boltzmann transport equation deterministically. It is also of interest to apply these analytical benchmarks to codes that produce the solution from simulated stochastic processes (Monte Carlo). The MCNP code is a multipurpose, coupled neutral, and charged-particle Monte Carlo transport code. Due to its accuracy and versatility, MCNP is rapidly becoming the Monte Carlo code of choice in the worldwide research and development community. The recent benchmark evaluations at Los Alamos National Laboratory include severalmore » critical and shielding measurements for various neutron applications but do not include formal comparisons with analytical solutions. The work discussed in the paper begins to address this need by applying MCNP to the searchlight problem in a semi-infinite medium. Also, this effort will extend the analytical benchmarking effort from personal computer platforms to a workstation environment.« less
  • The United States (US) Department of Energy Fissile Materials Disposition Program (FMDP) began studies for disposal of surplus weapons-grade plutonium (WG-Pu) as mixed uranium-plutonium oxide (@40X) fuel for commercial light-water reactors(LWRS). As a first step in this program, a test of the utilization of WG-Pu in a LWR environment is being conducted in an I-hole of the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). Initial radiation transport calculations of the test specimens were made at INEEL using the MCNP Monte Carlo radiation transport code to determine the linear heating rates in the fuel specimens.more » Unfortunately, the results of the calculations could not show the detailed high and low power-density spots in the specimens. Therefore, INEEL produced an MCNP source at the boundary of a rectangular parallelepiped enclosing the ATR I-hole, and Oak Ridge National Laboratory (ORNL) transformed this boundary source into a discrete -ordinates boundary source for the Three-dimensional Oak Ridge radiation Transport (TORT) code to pinpoint spatial detail. Agreement with average MCNP results were within 5%.« less