Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

Conference · Thu Jan 01 00:00:00 EST 1987

OSTI ID:5984681

Eyler, L L; Kim, D; Stover, R L; Beaver, T R

Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA); Pacific Northwest Lab., Richland, WA (USA)
• DOE Contract Number: AC06-76FF02170
• OSTI ID: 5984681
• Report Number(s): HEDL-SA-3645; PNL-SA-14511; CONF-871207-2; ON: DE87009973
• Resource Relation: Conference: Symposium on technology of glass, ceramic, or glass-ceramic to metal sealing, Boston, MA, USA, 13 Dec 1987; Other Information: Portions of this document are illegible in microfiche products
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
FFTF REACTOR
FUEL ASSEMBLIES
REACTOR COOLING SYSTEMS
THERMAL ANALYSIS
DESIGN
FINITE DIFFERENCE METHOD
FLOW MODELS
FUEL RODS
HEAT TRANSFER
HYDRAULICS
LITHIUM
NUMERICAL ANALYSIS
T CODES
ALKALI METALS
COMPUTER CODES
COOLING SYSTEMS
ELEMENTS
ENERGY SYSTEMS
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FLUID MECHANICS
FUEL ELEMENTS
ITERATIVE METHODS
LIQUID METAL COOLED REACTORS
MATHEMATICAL MODELS
MATHEMATICS
MECHANICS
METALS
NUMERICAL SOLUTION
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
TEST REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220300 - Nuclear Reactor Technology- Fuel Elements
220100 - Nuclear Reactor Technology- Theory & Calculation