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Title: MCNP-to-TORT radiation transport calculations for the Fissile Materials Disposition Program

Abstract

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 impact 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 amore » 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

Authors:
 [1]
  1. Oak Ridge National Lab., TN (United States)
Publication Date:
OSTI Identifier:
298343
Report Number(s):
CONF-981106-
Journal ID: TANSAO; ISSN 0003-018X; TRN: 99:001974
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 79; Conference: American Nuclear Society winter meeting, Washington, DC (United States), 15-19 Nov 1998; Other Information: PBD: 1998
Country of Publication:
United States
Language:
English
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; NUCLEAR WEAPONS DISMANTLEMENT; MIXED OXIDE FUELS; WATER COOLED REACTORS; RADIATION TRANSPORT; ATR REACTOR; M CODES; T CODES; GALLIUM; TEMPERATURE DISTRIBUTION

Citation Formats

Pace, III, J V. MCNP-to-TORT radiation transport calculations for the Fissile Materials Disposition Program. United States: N. p., 1998. Web.
Pace, III, J V. MCNP-to-TORT radiation transport calculations for the Fissile Materials Disposition Program. United States.
Pace, III, J V. Thu . "MCNP-to-TORT radiation transport calculations for the Fissile Materials Disposition Program". United States.
@article{osti_298343,
title = {MCNP-to-TORT radiation transport calculations for the Fissile Materials Disposition Program},
author = {Pace, III, J V},
abstractNote = {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 impact 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.},
doi = {},
url = {https://www.osti.gov/biblio/298343}, journal = {Transactions of the American Nuclear Society},
number = ,
volume = 79,
place = {United States},
year = {1998},
month = {12}
}