Abstracts and Available Papers Presented at the 1999 International RERTR Meeting
NEUTRONIC SAFETY PARAMETERS AND TRANSIENT ANALYSES FOR
POLAND’S MARIA RESEARCH REACTOR
M. M. Bretscher, N. A. Hanan, and
J. E. Matos
Argonne National Laboratory
Argonne, Illinois 60439-4841 USA
and
K. Andrzejewski and T. Kulikowska
Institute of Atomic Energy
Swierk, Poland
ABSTRACT
Reactor kinetic
parameters, reactivity feedback coefficients, and control rod reactivity worths
have been calculated for the MARIA Research Reactor (Swierk, Poland) for
M6-type fuel assemblies with 235U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for
family-dependent effective delayed neutron fractions, decay constants, and
prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were
determined for fuel Doppler coefficients, coolant (H2O) void and
temperature coefficients, and for in-core and ex-core beryllium temperature
coefficients. Total and differential
control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate
generic transients for fast and slow reactivity insertions with both HEU and
LEU fuels. The analyses show that the
HEU and LEU cores have very similar responses to these transients.