FUEL BURNUP STUDIES FOR A 225 Mwe ADVANCED SODIUM GRAPHITE REACTOR | Zendy

A. Aronson | Zendy

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FUEL BURNUP STUDIES FOR A 225 Mwe ADVANCED SODIUM GRAPHITE REACTOR

Author(s) -

A. Aronson

Publication year - 1960

Publication title -

osti oai (u.s. department of energy office of scientific and technical information)

Language(s) - English

Resource type - Reports

DOI - 10.2172/4168022

Subject(s) - burnup , criticality , nuclear engineering , reactivity (psychology) , graphite , neutron flux , homogeneous , nuclear reactor core , core (optical fiber) , materials science , radiochemistry , environmental science , nuclear physics , chemistry , physics , neutron , thermodynamics , engineering , medicine , alternative medicine , pathology , composite material

Reactivity and fuel burnup studies were performed for a 255 Mw(e) sodium- graphite reactor of the advanced calandria core type. This reactor is briefly described. Initial criticality calculations and flux distributions were obtained, using two-group theory for enrichments between 2.0 at.% U/sup 325/ and 4.0 at.% U235. A four-group burnup study was performed for enrichments between 2.5 at.% Uisup nd 3.25 at.% U/sup 235/. Core lifetime, changes in isotopic fuel composition, variations in radial power distribution, and fuel cross sec tions are presented. Reactivity during core lifetime was assumed to be controlled by the presence of a homogeneous poison which simulated the effects of control rcds. The results presentad are useful in determining initial enrichment selection in fuel programming and fuel cost studies. (auth

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