MCNP-to-TORT radiation transport calculations in support of mixed oxide fuels testing for the Fissile Materials Disposition Program
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.
- Research Organization:
- Oak Ridge National Lab., TN (United States)
- Sponsoring Organization:
- USDOE Office of Fissile Materials Disposition, Washington, DC (United States)
- DOE Contract Number:
- AC05-96OR22464
- OSTI ID:
- 650302
- Report Number(s):
- ORNL/CP--96969; CONF-980403--; ON: DE98003375
- Country of Publication:
- United States
- Language:
- English
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MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program
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