Criticality data and validation studies of plutonium-uranium nitrate solutions in annular geometry

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

OSTI ID:6996416

Smolen, G R ^[1]; Lloyd, R C ^[2]; Koyama, Tomozo ^[3]

---

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-Critical Mass Laboratory from 1985 to 1987 with mixed Pu + U nitrate solutions in annular geometry. The 25.4-cm-diam central region of the annual vessel contained various inserts, such as a bottle containing fissile solution and borated-concrete and cadmium-covered polyethylene annular inserts. The fissile solution concentrations ranged from 47 to 226 g Pu/[ell] with Pu/Pu + U ratios of 1.0, 0.5, and 0.2. The criticality data were used to validate two versions of the SCALE computer code system (SCALE-4 and SCALE-2). The analyses were performed with the 27-energy-group cross-section library, derived from the Evaluated Nuclear Data File B-Version IV. Computer models were prepared to accurately simulate all significant materials that would affect the system reactivity. The average calculated k[sub eff] for the 18 experiments was 1.008 ([sigma] = 0.006) with SCALE-4 and 1.004 ([sigma] = 0.006) with SCALE-2. Overall, the range of calculated k[sub eff]'s varied from 0.990 to 1.017. The results of the validation calculations indicate that the SCALE computer code system is capable of accurately modeling Pu + U nitrate solutions in annular geometry.

• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6996416
• 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
DEVELOPED COUNTRIES
FUEL REPROCESSING PLANTS
INTERNATIONAL COOPERATION
JAPAN
NATIONAL ORGANIZATIONS
NITRATES
NITROGEN COMPOUNDS
NUCLEAR DATA COLLECTIONS
NUCLEAR FACILITIES
OXYGEN COMPOUNDS
PLUTONIUM COMPOUNDS
PLUTONIUM NITRATES
S CODES
SAFETY ANALYSIS
TESTING
TRANSURANIUM COMPOUNDS
URANIUM COMPOUNDS
URANIUM NITRATES
US DOE
US ORGANIZATIONS
VALIDATION