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Title: MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program

Abstract

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. 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%.

Authors:
Publication Date:
Research Org.:
Oak Ridge National Lab., TN (US)
Sponsoring Org.:
USDOE Office of Science (US)
OSTI Identifier:
14348
Report Number(s):
ORNL/CP-104723
TRN: US0111006
DOE Contract Number:  
AC05-96OR22464
Resource Type:
Conference
Resource Relation:
Conference: 10th International Symposium on reactor Dosimetry, Osaka (JP), 09/12/1999--09/17/1999; Other Information: PBD: 1 Nov 1999
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 22 GENERAL STUDIES OF NUCLEAR REACTORS; MIXED OXIDE FUELS; RADIATION TRANSPORT; RADIOACTIVE WASTE DISPOSAL; WATER COOLED REACTORS; PLUTONIUM RECYCLE; ATR REACTOR; PERFORMANCE TESTING; MONTE CARLO METHOD; T CODES

Citation Formats

Pace, J V. MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program. United States: N. p., 1999. Web.
Pace, J V. MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program. United States.
Pace, J V. Mon . "MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program". United States. https://www.osti.gov/servlets/purl/14348.
@article{osti_14348,
title = {MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program},
author = {Pace, J V},
abstractNote = {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. 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%.},
doi = {},
url = {https://www.osti.gov/biblio/14348}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1999},
month = {11}
}

Conference:
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