Analytical Support for the ORR (Oak Ridge Research Reactor) Whole-Core LEU U₃Si₂-AL Fuel Demonstration

Conference · Mon Nov 03 00:00:00 EST 1986

OSTI ID:6624622

Bretscher, M. M. ^[1]

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1. Argonne National Laboratory (ANL), Argonne, IL (United States)

Analytical methods used to analyze neutronic data from the whole-core LEU fuel demonstration in the Oak Ridge Research Reactor are briefly discussed. Calculated eigenvalues corresponding to measured critical control rod positions are presented for each core used in the gradual transition from an all HEU to an all-LEU configuration. Some calculated and measured results, including β_eff/ℓ_p, are compared for HEU and LEU fresh fuel criticals. Finally, the perturbing influences of the six voided beam tubes on certain core parameters are examined. For reasons yet to be determined, differential shim rod worths are not well-calculated in partially burned cores.

• Research Organization: Argonne National Laboratory (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 6624622
• Report Number(s): CONF-861185-9; ON: DE87004710
• Resource Relation: Conference: Reduced Enrichment for Research and Test Reactors (RERTR) Program International Meeting, Gatlinburg, TN (United States), 3-6 Nov 1986; 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
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ORR REACTOR
REACTOR KINETICS
COMPUTER CALCULATIONS
CONTROL ELEMENTS
CONTROL ROD WORTHS
CRITICALITY
HIGHLY ENRICHED URANIUM
SLIGHTLY ENRICHED URANIUM
ACTINIDES
ELEMENTS
ENRICHED URANIUM
ENRICHED URANIUM REACTORS
ISOTOPE ENRICHED MATERIALS
KINETICS
MATERIALS
METALS
REACTOR COMPONENTS
REACTORS
TANK TYPE REACTORS
URANIUM
WATER COOLED REACTORS
WATER MODERATED REACTORS
Nuclear Criticality Safety Program (NCSP)
Neutronic Data
Highly Enriched Uranium (HEU)
Low Enriched Uranium (LEU)
Fresh Fuel Criticals
Shim Rod Worths
Transition Core
Equilibrium Core
High Uranium Density Fuel Element Development (HFED)
REBUS-3 Fuel Cycle Analyses Code
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220100 - Nuclear Reactor Technology- Theory & Calculation