Validation of the Scale Code System Using Criticality Data from Experiments Performed with Pu + U Solutions in Cylindrical and Slab Geometry
- Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
This paper outlines the results of a calculational study that was performed to validate two versions of the SCALE computer code system using data from critical experiments performed with mixed Pu + U aqueous solutions. The critical experiments were conducted in a 35-cm-diam cylinder and variable thickness slab tank. These experimental activities are part of a joint exchange program between the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The Consolidated Fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) manages the program for DOE. The experiments were conducted at the Battelle-Pacific Northwest Laboratories-Critical Mall Laboratory. Calculated multiplication factors have been examined for 36 critical experiments conducted in cylindrical and slab geometries with mixed Pu + U solutions. The results indicate that both the SCALE-2 and SCALE-4.0 computer code systems are capable of accurately modeling mixed fissile systems in simple geometry.
- Research Organization:
- Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- OSTI ID:
- 5419293
- Report Number(s):
- CONF-881011-; CODEN: TANSA; TRN: 89-029167
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 57; Conference: Joint meeting of the European Nuclear Society and the American Nuclear Society, Washington, DC (United States), 30 Oct - 4 Nov 1988; ISSN 0003-018X
- Publisher:
- American Nuclear Society
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
97 MATHEMATICS AND COMPUTING
CRITICALITY
S CODES
FUEL REPROCESSING PLANTS
SAFETY
PLUTONIUM NITRATES
VALIDATION
URANIUM NITRATES
AQUEOUS SOLUTIONS
BATTELLE PACIFIC NORTHWEST LABORATORIES
CYLINDERS
GEOMETRY
MULTIPLICATION FACTORS
NEUTRON REFLECTORS
ORNL
PNC
SLABS
US DOE
ACTINIDE COMPOUNDS
COMPUTER CODES
DISPERSIONS
JAPANESE ORGANIZATIONS
MATHEMATICS
MIXTURES
NATIONAL ORGANIZATIONS
NITRATES
NITROGEN COMPOUNDS
NUCLEAR FACILITIES
OXYGEN COMPOUNDS
PLUTONIUM COMPOUNDS
SOLUTIONS
TESTING
TRANSURANIUM COMPOUNDS
URANIUM COMPOUNDS
US AEC
US ERDA
US ORGANIZATIONS
Nuclear Criticality Safety Program (NCSP)
Calculational Study
Validation
Results
420203* - Engineering- Handling Equipment & Procedures
054000 - Nuclear Fuels- Health & Safety
050800 - Nuclear Fuels- Spent Fuels Reprocessing