A MONTE CARLO STUDY OF THERMAL UTILIZATION FACTOR AND DIFFUSION AREA: GAS- COOLED GRAPHITE-MODERATED LATTICES

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

OSTI ID:4262706

Schaefer, G W; Parkyn, D M

An attempt is made to explain the results from thermal flux in gas- cooled graphite-moderated uranium lattices experiments by solving the one-energy- group transport equation by the Monte Carlo method, using fundamental cross sections. The experiments of Campbell et al. and the theory of Amouyal et al. are compared with the Monte Carlo results. A survey of diffusion areas and asymmetries was carried out, using a number of mathematical models. The results are compared with theory and experiment. The survey is restricted to cylindrical fuel rods in circular channels and square pitches. (W.D.M.)

• Research Organization: English Electric Co., Ltd., Whetstone, Leics, Eng.
• NSA Number: NSA-13-007227
• OSTI ID: 4262706
• Report Number(s): A/CONF.15/P/310
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

ANALOG SYSTEMS
CONFIGURATION
COOLANT LOOPS
CROSS SECTIONS
CYLINDERS
DIFFUSION LENGTH
DISTRIBUTION
EQUATIONS
ERRORS
FUEL ELEMENTS
GAS COOLANT
GRAPHITE MODERATOR
GROUP THEORY
MATHEMATICS
MEASURED VALUES
MOCKUP
MONTE CARLO METHOD
NEUTRON FLUX
NUMERICALS
POWER PLANTS
REACTOR CORE
REACTORS
RODS
STANDARDS
THERMAL NEUTRONS
THERMAL UTILIZATION
TRANSPORT THEORY
URANIUM
ZONES