A Computer Study of Nuclear Characteristics of the University of Maryland Reactor
A Computer Study of Nuclear Characteristics of the University of Maryland Reactor
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Date
1967
Authors
Altomare, Philip M.
Advisor
Duffey, Dick
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Abstract
An analytical study of the University of Maryland
Reactor was performed using an IBM 7094 computer. The
nuclear analysis codes employed in the study were GAM-I,
for fast group cross-section evaluation, Tempest-II, for
thermal cross-sections, WANDA, for one-dimensional neutron
transport calculations and PDQ, for two-dimensional transport
calculations. A computer program, XS-1, for the calculation
of effective control rod cross-sections was written as part
of this study.
The method employed in the determination of fast energy
broad group cross-sections for non-fuel regions from that
normally used. The neutron leakage from the fuel was used as
a source of neutrons in lieu of the fission spectrum. This
method was felt to give a better estimate of the neutron flux
distribution in energy.
The analytical results compared favorably with experimental
data. The predicted K- effective was 0.993 versus 1.006 actual. The thermal neutron flux calculated and measured
at the center of fuel elements generally agreed within
10%. The calculated neutron flux above 3 Mev, which is more
difficult to match, was found to be low by approximately
30%. Control rod worth was predicted to be 5.6% Δρ for
the center shim rod and 2.7% Δρ for the outside shim rod.
Measured values are reported as 3.6% Δρ and 2.0% Δρ
respectively.