MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program [electronic resource].
- Washington, D.C. : United States. Dept. of Energy, 1999.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
- Physical Description:
- vp : digital, PDF file
- Additional Creators:
- Oak Ridge National Laboratory, United States. Department of Energy, and United States. Department of Energy. Office of Scientific and Technical Information
- Restrictions on Access:
- Free-to-read Unrestricted online access
- 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%.
- Report Numbers:
- E 1.99:ornl/cp-104723
- Published through SciTech Connect.
10th International Symposium on reactor Dosimetry, Osaka (JP), 09/12/1999--09/17/1999.
- Funding Information:
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