Program - Brookhaven National Laboratory

safety cases from the International Criticality Safety Benchmark Evaluation Project, ranging from lowenriched
uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uraniumplutonium,
metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks
were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si,
Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments
at Lawrence Livermore National Laboratory (for 6 Li, 7 Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O,
concrete, polyethylene and teflon). The new functionality in MCNP to calculate the effective delayed
neutron fraction was tested by comparison with more than thirty measurements in widely varying systems.
Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil),
both with a thermal spectrum, and two cores in Masurca (France) and three cores in the Fast Critical
Assembly (FCA, Japan), all with fast spectra.
RE 3 11:20 AM
Bayesian Statistical Analysis Applied to NAA Data for Neutron Flux Spectrum
Determination
Davide Chiesa, Massimiliano Clemenza, Monica Sisti
Milano Bicocca University, Italy
Luca Pattavina, Ezio Previtali
Istituto Nazionale di Fisica Nucleare (INFN), Milano Bicocca, Italy
Andrea Borio di Tigliole, Andrea Salvini
Laboratorio di Energia Nucleare Applicata (LENA), Pavia University, Italy
Antonio Cammi
Polytechnic University of Milan, Italy
In this paper, we present a statistical methodology to analyze the neutron activation data and to evaluate
the neutron flux and its energetic distribution. The neutron activation analysis (NAA) technique was used
to perform an absolute measurement of the neutron flux. Many samples containing a known amount of
parent nuclei were irradiated at the TRIGA Mark II reactor of Pavia University (Italy) and the activation
rate of a large number of isotopes was measured through gamma-ray spectroscopy with very low background
HPGe detectors. In order to have a precise determination of the activation rate, the measurements were
repeated on different HPGe detectors and Monte Carlo codes based on GEANT4 were developed to evaluate
the gamma detection efficiency for every radioisotope of interest. Then the activation data of the different
isotopes were combined to evaluate the energy spectrum of the neutron flux. For this purpose a system
of linear equations, containing the group cross section data and the experimental results of the activation
rate, should be solved. Since the coefficients and the parameters of this system are affected by experimental
uncertainties, a rigorous statistical approach is fundamental to get the correct physics results. The Bayesian
statistical approach allows to solve this kind of problems by including the uncertainties of the coefficients
and the a priori information about the flux. A program for the analysis of Bayesian hierarchical models,
based on Markov Chain Monte Carlo (MCMC) simulations, was used to describe a statistical model of the
problem and to determine the flux energy group distributions. This methodology has been tested using
the activation data acquired in three different irradiation facilities of the TRIGA reactor. The results
presented in this paper show that this technique of analysis is promising, since it allows the evaluation of
the flux energy groups and their uncertainties with great precision. The dependence of the results on the
prior distribution choice and on the group cross section data was investigated, confirming the reliability of
the analysis. Finally, the correlations between the different energy groups were evaluated, underlining the
degree of information added by the experiment.
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