Predrag Marinkovic

A study on criticality of coupled fast-thermal core HERBE at RB reactor

Abstract The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim to improve experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and application of available fuel restricted design flexibility of the coupled system. Thus the following core is considered optimal in the sense of the foregoing constraints: the central fast core of natural uranium is surrounded by neutron filter zone (cadmium and natural uranium) and converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. Thermal core is formed of RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by computer code package for the design of thermal and coupled fast-thermal reactors recently developed in IBK Nuclear Engineering Laboratory. Good agreement between computations and experimental data was achieved.

The effect of the neutron spectra unfolding method on the fast neutron dose determination

Abstract Based on Shanons information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms.

The effects of shielding around the source on neutron energy spectra

The knowledge of neutron energy spectra contributes to unambiguous identification of neutron sources in the fields of nuclear safeguards and nuclear non-proliferation. Since a real scenario situation includes the presence of shielding around the source, we have investigated the influence of the potential shielding surrounding the source on the shape of energy spectra for a few neutron sources. We have applied the maximum-likelihood, expectation–maximisation (MLEM) method with one-step-late (OSL) algorithm for neutron spectra unfolding. The pulse height distributions used in the unfolding procedures were simulated with the high accuracy by using the MCNP-PoliMi code based on the Monte Carlo method. A possibility to identify the shielded neutron sources by using the unfolding method was examined with two continuous-in-energy sources, such as 252Cf and 241Am–Be in source-shielding configurations with lead (Pb) and polyethylene (PE) blocks. The results of calculations have shown that the identification of 252Cf and 241Am–Be sources with 2.5 cm of Pb and PE shield can be achieved successfully by using the MLEM method with the OSL algorithm. However, the unfolded results for 252Cf and 241Am–Be sources with 10 cm of PE shield significantly deviate from the reference spectra and the sources cannot be correctly identified on the basis of their unfolded energy spectra.

Identification of Neutron Sources by Spectral Analysis of Pulse Height Distributions

This paper proposes a neutron source identification method based on the spectral analysis of neutron pulse height distributions obtained with liquid scintillation detectors. The fact that shielded and unshielded neutron sources have clearly defined spectral components with specific locations and intensities offers the possibility of identifying the sources based on spectral features alone, without having to unfold the energy spectra. Analysis of simulated and experimental data confirms that this new identification method is promising, and that good resolution power can be achieved.

A 3D point-kernel multiple scatter model for parallel-beam SPECT based on a gamma-ray buildup factor

A three-dimensional (3D) point-kernel multiple scatter model for point spread function (PSF) determination in parallel-beam single-photon emission computed tomography (SPECT), based on a dose gamma-ray buildup factor, is proposed. This model embraces nonuniform attenuation in a voxelized object of imaging (patient body) and multiple scattering that is treated as in the point-kernel integration gamma-ray shielding problems. First-order Compton scattering is done by means of the Klein-Nishina formula, but the multiple scattering is accounted for by making use of a dose buildup factor. An asset of the present model is the possibility of generating a complete two-dimensional (2D) PSF that can be used for 3D SPECT reconstruction by means of iterative algorithms. The proposed model is convenient in those situations where more exact techniques are not economical. For the proposed models testing purpose calculations (for the point source in a nonuniform scattering object for parallel beam collimator geometry), the multiple-order scatter PSF generated by means of the proposed model matched well with those using Monte Carlo (MC) simulations. Discrepancies are observed only at the exponential tails mostly due to the high statistic uncertainty of MC simulations in this area, but not because of the inappropriateness of the model.

Radiography simulation based on point-kernel model and dose buildup factors

Three-dimensional point-kernel multiple scatter model for radiography simulation, based on dose X-ray buildup factors, is proposed and validated to Monte Carlo simulation. This model embraces nonuniform attenuation in object of imaging (patient body tissue). Photon multiple scattering is treated as in the point-kernel integration gamma ray shielding problems via scatter voxels. First order Compton scattering is described by means of Klein-Nishina formula. Photon multiple scattering is accounted by using dose buildup factors. The proposed model is convenient in those situations where more exact techniques, like Monte Carlo, are not time consuming acceptable.

Kernel-Integration Scatter Model for the Characterization of Nuclear Waste Drums by Gamma Emission Tomography

The objective of this study is the determination of the distribution of the activity of radio-elements contained in radioactive waste packages by means of single photon emission computed tomography (SPECT). A three- dimensional (3D) projector simulator for a parallel hole collimator and NaI(Tl) scintillator, based on the point kernel integration method, is proposed. The model takes into account the attenuation and scatter of gamma rays. Primary scattered photons are treated by the Compton process and Klein-Nishina formula, and the multiple scattering is accounted for by means of the dose buildup factor normally used in shielding problems. An advantage of the proposed model is that it offers the possibility of generating a full two-dimensional point-spread function (PSF) that can be used for 3D reconstruction. The developed model is convenient in those situations where more exact techniques (such as Monte Carlo simulation) are not economical. The model has been evaluated for the package having homogenous density with the Cs-137 point source inside.

Toward utilization of MCNP5 particle track output file for simulation problems in photon spectrometry

Abstract Pulse height distribution (PHD) registered by a spectrometer is influenced by various physical phenomena such as photon interactions as well as disturbance produced by the electronic circuits inside the spectrometer. Therefore, spectrometry measurements of gamma and X-ray radiation inaccurately represent primary spectra. In order to overcome spectrum disruption, spectrum unfolding has to be applied. One of the common tools used in the unfolding process is Monte Carlo simulation of spectrometer response to monochromatic photons. The purpose of this work is to develop a new method for simulating CdTe semiconductor spectrometer response to monochromatic photons that can be further used for the spectrum unfolding procedure. The method is based upon post-processing of the particle track (PTRAC) output file generated by the MCNP5 program. In addition to the spectrometry output, this method provides information for each specific photon interaction inside the spectrometer active volume, which is required when taking into account spectrometer charge collection. The PTRAC generated detector response and the measured spectrum were in good agreement. The results obtained showed that this method can be used to generate precise response functions of gamma and X-ray spectrometers.

Radiography Simulation Based on Exposure Buildup Factors for Multilayer Structures

Monte Carlo techniques were usually used to study the effect of scattered photons on a radiographic X-ray image. Such approach is accurate, but computer time consuming. On the other hand, the exposure buildup factors can be used as approximate and efficient assessment to account for the scattering of X-rays. This method uses the known radiography parameters to find the resulting detector exposure due to both scattered and un-collided photons. A model for radiography simulation, based on X-ray dose buildup factor, is proposed. This model includes non-uniform attenuation in voxelized object of imaging (patient body tissue). Composition of patient body is considered as a multi-layer structure. Various empirical formulas exist for multi-layer structure calculations and they all calculate multi-layer buildup factors by combining single-layer buildup factors. The proposed model is convenient in cases when more exact techniques (like Monte Carlo) are not economical.

Approach to unfolding neutron spectra measured by a 3He semiconductor detector

In measurements of fast neutron spectra by a 3He semiconductor detector, the unfolding method is not usually required. The unfolding method based on principle of maximum likelihood that incorporates the Gaussian approximation of counting statistics is developed and implemented in the MLMHE3 numerical code for application in fast neutron spectrometry by 3He semiconductor detectors. The derived likelihood equations have been solved using method of the singular value decomposition of the response matrix. For this inverse problem, the detector responses were generated by the Monte Carlo technique.