Argonne National Laboratory
RERTR
Reduced Enrichment for Research and Test Reactors
Nuclear Engineering Division at Argonne
U.S. Department of Energy

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Abstracts and Available Papers Presented at the
1999 International RERTR Meeting

NEUTRONIC SAFETY PARAMETERS AND TRANSIENT ANALYSES FOR POLANDíS 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.


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Contact:
Manuel M. Bretscher
Physicist
Argonne National Laboratory Ė 362
9700 South Cass Avenue
Argonne, IL 60439 USA

Phone:† (630) 252-8616
Fax:††††† (630) 252-5161

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Last modified on July 29, 2008 11:33 +0200