Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

Conference · Mon Sep 27 00:00:00 EDT 1999

OSTI ID:750510

Bretscher, M M; Hanan, N A; Matos, J E; Andrzejewski, K; Kulikowska, T

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 {sup 235}U 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 (H{sub 2}O) 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.

• Research Organization: Argonne National Lab., IL (US)
• Sponsoring Organization: US Department of Energy (US)
• DOE Contract Number: W-31-109-ENG-38
• OSTI ID: 750510
• Report Number(s): ANL/TD/CP-100104; TRN: US0204637
• Resource Relation: Conference: 1999 International Meeting on Reduced Enrichment for Research and Test Reactors, Budapest (HU), 10/03/1999--10/08/1999; Other Information: PBD: 27 Sep 1999
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS
CONTROL ELEMENTS
CONTROL ROD WORTHS
DELAYED NEUTRON FRACTION
DOPPLER COEFFICIENT
FUEL ASSEMBLIES
POLAND
PROMPT NEUTRONS
REACTIVITY INSERTIONS
REACTIVITY WORTHS
REACTOR KINETICS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SCRAM RODS
TEMPERATURE COEFFICIENT
TRANSIENTS