Criticality data and validation studies of mixed-oxide fuel pin arrays in Pu+U+Gd nitrate

Journal Article · Thu Sep 01 04:00:00 EDT 1994 · Nuclear Technology; (United States)

OSTI ID:6903597

Smolen, G R ^[1]; Lloyd, R C ^[2]; Matsumoto, Tadakuni ^[3]

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1. Oak Ridge National Lab., TN (United States). Computing and Telecommunications Div.
2. Pacific Northwest Lab., Richland, WA (United States)
3. Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

Critical experiments were performed at the Pacific Northwest Laboratory's Critical Mass Laboratory in 1987 and 1988 with a heterogeneous array of mixed-oxide (MOX) fuel pins immersed in mixed plutonium-uranium nitrate solutions. The 996 fuel pins, on a 1.40-cm-square pitch, were configured in a cylindrical array. The solution heavy metal concentrations ranged from 4 to 468 g/[ell] and had a Pu/Pu + U ratio of 0.2. Critical experiments were also performed with gadolinium added to the fissile solution. These experiments were designed to simulate conditions in a MOX fuel dissolver, where fuel lumps are moderated by aqueous solutions containing fissile nuclides, with and without a soluble neutron poison. For the experimental conditions examined, it was determined that the critical size of the system increased as the heavy metal concentration increased. The criticality data were used to validate two versions of the scale computer code system and the 27-energy-group cross-section library, derived from the Evaluated Nuclear Data File B Version IV. The calculations results indicate that SCALE-2 has some difficulty in modeling these systems. Modifications in SCALE-4 have led to more accurate K[sub eff] results.

• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6903597
• Journal Information: Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 107:3; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

Related Subjects

050800 -- Nuclear Fuels-- Spent Fuels Reprocessing
054000* -- Nuclear Fuels-- Health & Safety
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
420203 -- Engineering-- Handling Equipment & Procedures
ACTINIDE COMPOUNDS
ASIA
COMPUTER CODES
COOPERATION
CRITICALITY
DESIGN
DEVELOPED COUNTRIES
ELEMENTS
FUEL ELEMENTS
FUEL PINS
FUEL REPROCESSING PLANTS
FUELS
GADOLINIUM
INTERNATIONAL COOPERATION
JAPAN
METALS
MIXED OXIDE FUELS
MODERATORS
NATIONAL ORGANIZATIONS
NEUTRON ABSORBERS
NITRATES
NITROGEN COMPOUNDS
NUCLEAR DATA COLLECTIONS
NUCLEAR FACILITIES
OXYGEN COMPOUNDS
PLUTONIUM COMPOUNDS
PLUTONIUM NITRATES
RARE EARTHS
REACTOR COMPONENTS
S CODES
SAFETY ANALYSIS
SOLID FUELS
TESTING
TRANSURANIUM COMPOUNDS
URANIUM COMPOUNDS
URANIUM NITRATES
US DOE
US ORGANIZATIONS
VALIDATION