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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 US (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). Currently MOX fuel is used commercially in a number of foreign countries, but is not in the US. Most of the experience is with reactor grade plutonium (RG-Pu) in MOX fuel. Therefore, to use WG-Pu in MOX fuel, one must demonstrate that the experience with RG-Pu is relevant. As a first step in this program, the utilization of WG-Pu in a LWR environment must be demonstrated. To accomplish this, a test is to be conducted to investigate some of the unresolved issues. The initial tests will be made 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. These calculations were made 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. However, a discrete ordinates radiation transport codemore » could pinpoint these spatial details. Therefore, INEEL was tasked with producing a MCNP source at the boundary of a rectangular parallelepiped enclosing the ATR I-hole, and Oak Ridge National Laboratory was tasked with transforming this boundary source into a discrete ordinates boundary source for the Three-dimensional Oak Ridge radiation Transport (TORT) code. Thus, the TORT results not only complemented, but also were in agreement with the MCNP results.« less

Authors:
Publication Date:
Research Org.:
Oak Ridge National Lab., TN (United States)
Sponsoring Org.:
USDOE Office of Fissile Materials Disposition, Washington, DC (United States)
OSTI Identifier:
650302
Report Number(s):
ORNL/CP-96969; CONF-980403-
ON: DE98003375; TRN: 98:010044
DOE Contract Number:  
AC05-96OR22464
Resource Type:
Conference
Resource Relation:
Conference: Radiation protection and shielding topical meeting: technologies for the new century, Nashville, TN (United States), 19-23 Apr 1998; Other Information: PBD: Apr 1998
Country of Publication:
United States
Language:
English
Subject:
05 NUCLEAR FUELS; 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; WATER COOLED REACTORS; NUCLEAR MATERIALS MANAGEMENT; PLUTONIUM; MIXED OXIDE FUELS; PERFORMANCE TESTING; ATR REACTOR; COMPUTERIZED SIMULATION

Citation Formats

Pace, III, 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., 1998. Web.
Pace, III, J V. MCNP-to-TORT radiation transport calculations in support of mixed oxide fuels testing for the Fissile Materials Disposition Program. United States.
Pace, III, J V. Wed . "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/650302.
@article{osti_650302,
title = {MCNP-to-TORT radiation transport calculations in support of mixed oxide fuels testing for the Fissile Materials Disposition Program},
author = {Pace, III, J V},
abstractNote = {The US (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). Currently MOX fuel is used commercially in a number of foreign countries, but is not in the US. Most of the experience is with reactor grade plutonium (RG-Pu) in MOX fuel. Therefore, to use WG-Pu in MOX fuel, one must demonstrate that the experience with RG-Pu is relevant. As a first step in this program, the utilization of WG-Pu in a LWR environment must be demonstrated. To accomplish this, a test is to be conducted to investigate some of the unresolved issues. The initial tests will be made 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. These calculations were made 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. However, a discrete ordinates radiation transport code could pinpoint these spatial details. Therefore, INEEL was tasked with producing a MCNP source at the boundary of a rectangular parallelepiped enclosing the ATR I-hole, and Oak Ridge National Laboratory was tasked with transforming this boundary source into a discrete ordinates boundary source for the Three-dimensional Oak Ridge radiation Transport (TORT) code. Thus, the TORT results not only complemented, but also were in agreement with the MCNP results.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1998},
month = {4}
}

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