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

PDF Version Also Available for Download.

Description

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. ... continued below

Physical Description

vp.

Creation Information

Pace, J.V. November 1, 1999.

Context

This article is part of the collection entitled: Office of Scientific & Technical Information Technical Reports and was provided by UNT Libraries Government Documents Department to Digital Library, a digital repository hosted by the UNT Libraries. More information about this article can be viewed below.

Who

People and organizations associated with either the creation of this article or its content.

Author

Sponsor

Publisher

Provided By

UNT Libraries Government Documents Department

Serving as both a federal and a state depository library, the UNT Libraries Government Documents Department maintains millions of items in a variety of formats. The department is a member of the FDLP Content Partnerships Program and an Affiliated Archive of the National Archives.

Contact Us

What

Descriptive information to help identify this article. Follow the links below to find similar items on the Digital Library.

Description

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%.

Physical Description

vp.

Notes

INIS; OSTI as DE00014348

Source

  • 10th International Symposium on reactor Dosimetry, Osaka (JP), 09/12/1999--09/17/1999

Language

Item Type

Identifier

Unique identifying numbers for this article in the Digital Library or other systems.

  • Report No.: ORNL/CP-104723
  • Grant Number: AC05-96OR22464
  • Office of Scientific & Technical Information Report Number: 14348
  • Archival Resource Key: ark:/67531/metadc620469

Collections

This article is part of the following collection of related materials.

Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

What responsibilities do I have when using this article?

When

Dates and time periods associated with this article.

Creation Date

  • November 1, 1999

Added to The UNT Digital Library

  • June 16, 2015, 7:43 a.m.

Description Last Updated

  • Jan. 19, 2016, 7:43 p.m.

Usage Statistics

When was this article last used?

Yesterday: 0
Past 30 days: 0
Total Uses: 9

Interact With This Article

Here are some suggestions for what to do next.

Start Reading

PDF Version Also Available for Download.

International Image Interoperability Framework

IIF Logo

We support the IIIF Presentation API

Pace, J.V. MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program, article, November 1, 1999; Tennessee. (digital.library.unt.edu/ark:/67531/metadc620469/: accessed October 19, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.