AN INTERMEDIATE REACTOR FOR OBTAINING HIGH INTENSITY NEUTRON FLUXES

Technical Report · Sat Oct 31 04:00:00 EST 1959

OSTI ID:4279255

Feinberg, S M; Vorobiev, E D

An intermediate research reactor designed for obtaining high thermal and fast neutron fluxes is described. The reactor consists of a core with plate fuel elements containing highly enriched uranium. The fuel elements are cooled by water flowing through the gaps between the plates. The core is surrounded by a reflector of beryllium oxide. In the center of the core there is a water thermal column where the maximum thermal neutron flux is produced. The ratio of the maximum neutron flux to heat output characterizing the performance of the research reactor is several times greater than that for thermal reactors. Theoretical premises for the use of intermediate reactors are given and their advantages are shown. The results of experiments with critical assemblies carried out with the aim of studying uranium--water intermediate reactor physics are discussed. (auth)

• Research Organization: Academy of Sciences, U.S.S.R.
• NSA Number: NSA-13-007177
• OSTI ID: 4279255
• Report Number(s): A/CONF.15/P/2142
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

BERYLLIUM OXIDES
CONFIGURATION
COOLANT LOOPS
CRITICAL ASSEMBLIES
CRITICALITY
EFFICIENCY
EPITHERMAL NEUTRONS
EQUATIONS
FAST NEUTRONS
FUEL ELEMENTS
HEAT TRANSFER
MEASURED VALUES
MODERATORS
MULTIPLICATION FACTORS
NEUTRON FLUX
PLATES
REACTOR CORE
REACTORS
REFLECTORS
RESEARCH REACTORS
THERMAL COLUMN
THERMAL NEUTRONS
URANIUM
URANIUM 235
USSR
WATER
WATER COOLANT
WATER MODERATOR
ZONES