HFIR PRELIMINARY PHYSICS REPORT

Technical Report · Wed Oct 19 00:00:00 EDT 1960

DOI:https://doi.org/10.2172/4123862· OSTI ID:4123862

Cheverton, R. D.

High Flux Isotope Reactor physics studies were conducted to establish a core design that would permit the attainment of about 5 x 10/sup 15/ neutrons/cm/ sup 2/-sec unperturbed thermal flux at a power level of 100 Mw. The proposed reactor is a cylindrical flux-trap reactor, having a lightwater island surrounded by a fuel annulus containing fully enriched uranium, water, and aluminum. A vertical-plate control system separates the fuel region from the watercooled beryllium side reflector. The island diameter is 14 cm; the fuel-annulus outer diameter is 38 cm; and the active core length is 45.7 cm. It is estimated that this core with enough fuel for 10 days of operation at 100 Mw will produce a maximum unperturbed thermal flux of 6.6 x 10/sup 15/ neutrons/cm/sup 2/-sec and an average thermal flux of 3.7 x 10/sup 15/ in a 200-g sample of Pu/sup 242/. The corresponding maximum heat flux was estimated to be 1.6 x 10/sup 6/ Btu/hr-ft/ sup 2/ for a fuel element containing 0.050-in.-thick plates and 0.050-in.thick coolant channels. (auth)

• Research Organization: Oak Ridge National Lab., Tenn.
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-15-001081
• OSTI ID: 4123862
• Report Number(s): ORNL-3006
• Country of Publication: United States
• Language: English

Related Subjects

ALUMINUM
BERYLLIUM
CONTROL ELEMENTS
COOLANT LOOPS
ENRICHMENT
FUEL ELEMENTS
HEAT TRANSFER
HFIR
ISOTOPES
NEUTRON FLUX
Nuclear Criticality Safety Program (NCSP)
PLATES
PLUTONIUM 242
PRODUCTION
REACTOR CORE
REACTOR TECHNOLOGY
REACTORS
REFLECTORS
RESEARCH REACTORS
THERMAL NEUTRONS
TRAPS
URANIUM
WATER
WATER COOLANT
ZONES