Three-dimensional radiation dose mapping with the TORT computer code

Conference · Tue Jan 01 00:00:00 EST 1991 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5864800

Slater, C O; Pace, III, J V; Childs, R L; Haire, M J ^[1]; Koyama, T ^[2]

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1. Oak Ridge National Lab., TN (United States)
2. Power Reactor and Nuclear Fuels Development Corp., Ibaraki (Japan)

The Consolidated Fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) has performed radiation shielding studies in support of various facility designs for many years. Computer codes employing the point-kernel method have been used, and the accuracy of these codes is within acceptable limits. However, to further improve the accuracy and to calculate dose at a larger number of locations, a higher order method is desired, even for analyses performed in the early stages of facility design. Consequently, the three-dimensional discrete ordinates transport code TORT, developed at ORNL in the mid-1980s, was selected to examine in detail the dose received at equipment locations. The capabilities of the code have been previously reported. Recently, the Power Reactor and Nuclear Fuel Development Corporation in Japan and the US Department of Energy have used the TORT code as part of a collaborative agreement to jointly develop breeder reactor fuel reprocessing technology. In particular, CFRP used the TORT code to estimate radiation dose levels within the main process cell for a conceptual plant design and to establish process equipment lifetimes. The results reported in this paper are for a conceptual plant design that included the mechanical head and (i.e., the disassembly and shear machines), solvent extraction equipment, and miscellaneous process support equipment.

• OSTI ID: 5864800
• Report Number(s): CONF-910603-; CODEN: TANSA
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 63; Conference: Annual meeting of the American Nuclear Society (ANS), Orlando, FL (United States), 2-6 Jun 1991; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
61 RADIATION PROTECTION AND DOSIMETRY
FUEL REPROCESSING PLANTS
RADIATION PROTECTION
SHIELDING
T CODES
ACCURACY
ANL
COMPUTER CALCULATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
COORDINATED RESEARCH PROGRAMS
DISCRETE ORDINATE METHOD
FUEL ASSEMBLIES
NEUTRON ABSORBERS
NEUTRON TRANSPORT
ORNL
PNC
RADIATION DOSES
REPROCESSING
SAFETY
SOLVENT EXTRACTION
SPENT FUELS
THREE-DIMENSIONAL CALCULATIONS
US DOE
VALIDATION
VERIFICATION
DOSES
ENERGY SOURCES
EXTRACTION
FUELS
JAPANESE ORGANIZATIONS
MATERIALS
NATIONAL ORGANIZATIONS
NEUTRAL-PARTICLE TRANSPORT
NUCLEAR FACILITIES
NUCLEAR FUELS
RADIATION TRANSPORT
REACTOR MATERIALS
RESEARCH PROGRAMS
SEPARATION PROCESSES
SIMULATION
TESTING
US AEC
US ERDA
US ORGANIZATIONS
054000* - Nuclear Fuels- Health & Safety
050800 - Nuclear Fuels- Spent Fuels Reprocessing
560190 - Radiation Protection Standards- (1992-)