RRFM 2009 Transactions - European Nuclear Society

ANALYSES OF IRT, SOFIA LEU CORE:
STEADY-STATE THERMAL-HYDRAULIC AND ACCIDENT
ANALYSES
S.I. BELOUSOV, T.G. APOSTOLOV
Institute for Nuclear Research and Nuclear Energy of Bulgarian Academy of Science
Tsarigradsko 72, 1784 Sofia, Bulgaria
N.A. HANAN, J.E. MATOS
RERTR Program, Argonne National Laboratory
Argonne, IL 60439-4815 USA
ABSTRACT
The initial LEU (IRT-4M fuel assemblies, 19.75% 235 U) core of the new IRT, Sofia
research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE)
of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly analyzed with the
RERTR Program at Argonne National Laboratory (ANL) to evaluate its
characteristics important to safety. Analyses are carried out for the fuel properties
presented in the actual fuel Catalogue Description specified for IRT, Sofia research
reactor. The initial configuration using 16 fuel assemblies (four 8-tube and twelve
6-tube fuel assemblies) is analyzed at a critical core state preferable for BNCT
tube operation.
Results of detailed steady-state thermal-hydraulic calculations and transients for
this initial core configuration are presented in this paper. These results show that
one pump in the primary circuit is sufficient for safe operation at 200, 500 and 1000
kW reactor power levels and that safety is maintained for all transients.
1. Introduction
A joint study concerning IRT, Sofia research reactor (RR) between INRNE and the RERTR
Program at ANL was initiated in 2002. The initial steps studies [1-4] were mainly focused on
neutronics properties significant for reactor application and safety analyses. The results of
further neutronics thermal hydraulic and accident analyses significant for safety assessment
of the LEU core for the critical state preferable for Boron Neutron Capture Therapy (BNCT)
beam tube operation were presented too [5, 6]. The BNCT activity is considered nowadays
as one of the most important for future IRT, Sofia application. However these results were
based on the preliminary data for the LEU fuel assemblies (FA) [7, 8] that differed from the
actual fuel data [9] which became available later. Presented here are results of the steadystate
and transient thermal hydraulic calculations using the actual LEU fuel assemblies’ data.
The actual fuel data that are most significant for the safety assessment analysis and differ
from the previously used include: the FA maximum inlet coolant temperature; the maximum
admissible temperature on the fuel element (FE) surface; and onset of nucleate boiling ratio
(ONBR). Moreover the hot channel factors (HCF) impact is not considered in the actual fuel
passport. The actual fuel data include the temperature of ONB (118ºC) that was not
presented in the previously used data. The actual catalogue description also defines the
fraction of the reactor power generated in the FE equal to 95% whereas the value of 94%
was used in our previous calculations. In the safety assessment it is important to remember
that increasing this fraction leads to corresponding increase of the peak power density in the
core; this needs to be included in the thermal-hydraulics analyses (steady-state and
transients).
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