Program - Brookhaven National Laboratory

For experimental determination of the thermal flux in RPV and pin power density in fuel the VVER-1000
Mock-up on reactor LR-0 was utilized. The thermal neutron flux distribution was determined by means
of 3He reaction rate. The 3He proportional counter 6NH12.5 Canberra (active diameter 9mm, active
length 125mm, 3He pressure 8 bar) was used for this purpose. The pin power density was measured by
means of fission rate density which was determined using fission product activity. For this measurement,
semiconductor gamma spectrometry with high energy resolution (approximately 2 keV at Eγ=1333 keV)
and multichannel analyzer DSA2000 (Canberra) were employed. The high purity germanium coaxial
detector (Ortec, relative efficiency 70%) was placed in a thick Pb cylindrical shield with 1x2 cm collimator.
The calculations were performed using MCNP code (version MCNPX 2.6.0.). Various sets of nuclear data
libraries were used in the calculation.
PR 45
Sensitivity and Uncertainty Analysis of the GFR MOX Fuel Subassembly
J. Lüley, B. Vrban, S. Cerba, J. Hascik, V. Nečas, Institute of Nuclear and Physical Engineering, Slovak
University of Technology in Bratislava, Ilkovičova 3, 812 19 Bratislava, Slovakia. S. Pelloni, Laboratory
for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI, Switzerland.
In recent years nuclear energy has seen a renaissance giving the opportunity for innovative conceptual
designs. For their deployment nuclear data definitely plays a role, but adequate validations and experimental
data bases are so far largely missing [1]. A correct assessment of the uncertainty of the analytical
parameters related to the reactor core performance resulting from uncertainties in the nuclear data is an
important issue for designing a reactor, and it is certainly useful for identifying potential sources of computational
biases. Thereby, sensitivity analyses provide a unique insight into the system performance being
applicable to propagate the uncertainties in the cross-section data to uncertainties of the system response
[2]. This study deals with sensitivity and uncertainty analyses and a benchmark similarity assessment of
the MOX fuel subassembly designed for the Gas Cooled Fast Reactor (GFR) as a representative material
of the core, serving to identify the main contributors to the calculation bias.
Corresponding author: Vladimir Necas
[1] V. Jagannathan, U. Pal, R. Karthikeyan, A. Srivastava and S. A. Khan, Sensitivity to Nuclear Data
Libraries in the Physics Core Characteristics of Conceptual Thorium Breeders, Journal of the Korean
Physical Society, Vol. 59, No. 2, August 2011, pp. 1361-1364 [2] B.T. Rearden et al., Sensitivity and
Uncertainty Analysis Capabilities and Data in Scale, Nuclear Technology, Vol. 174, May 2011
PR 46
Results of Calculations of Criticality Parameters of Liquid Metal-Cooled Fast Benchmark
Systems with the Use of ABBN Data Set
Vladimir Koscheev, Gleb Lomakov, Gennady Manturov, Anton Peregudov, Mikhael Semenov, Anatoly
Tsibulya, Institute for Physics and Power Engineering Bondarenko Square 1, Obninsk 244033, Kaluga
Region, Russia.
The paper contains results of performed calculations of benchmark reactor-physics configurations from International
Reactor Physics Experiment Evaluation Project (IRPHEP). Calculations of k-effective, spectral
indices, sodium void coefficients, efficiency of control rods and fission rates distributions were compared
with the experimental data. The set of experiments covers various types of fuel used in fast reactors -
uranium, MOX and metal fuel. The calculations were performed using Monte-Carlo codes MMKKENO
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