MCNP-PoliMi Code Research Articles

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.

A New Scintillator for Fast Neutron Detection: Single-Crystal ${\rm CeCl}_{3}({\rm CH}_{3}{\rm OH})_{4}$

A new scintillator material for the detection of fast neutrons that consists of a methanol adduct of cerium trichloride with the composition CeCl <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> (CH <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> OH) <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</sub> has been characterized using 14.1 MeV neutrons from a deuterium-tritium neutron generator and fast neutrons from a bare instrumented <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">252</sup> Cf source. The timing resolution of the scintillator for fast neutrons was found to be 1~ ns. Neutron interactions in the CeCl <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> (CH <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> OH) <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</sub> composition were simulated using the MCNP-PoliMi code. These simulations indicate that proton recoils account for most of the deposited energy in CeCl <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> (CH <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> OH) <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</sub> . The crystalline molecular adduct CeCl <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> (CH <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> OH) <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</sub> represents a rare-earth metal-organic scintillator that can be applied to both fast neutron and gamma-ray detection.

Pulse-height distributions of neutron and gamma rays from plutonium-oxide samples

We present new results on neutron and gamma-ray pulse-height distributions (PHDs) measured with liquid scintillators from five plutonium-oxide samples of varying mass and burnup and a 252Cf isotopic source. We show that the analysis of the pulse-height distributions can be used to easily distinguish the fissile material (plutonium oxide) from the 252Cf source. Moreover, the slope of the measured pulse-height distributions can be analyzed to distinguish the burnup of the samples, independent of their masses. Finally, the measurement scenarios are modeled using the MCNP-PoliMi code and good agreement is observed between the measured and the simulated pulse-height distributions.

Neutron and gamma-ray cross-correlation measurements of plutonium oxide powder

For the first time, measurements of the time-dependent cross-correlation distributions of plutonium oxide have been made separately for neutrons and gamma rays. Six EJ-309 liquid scintillation detectors with a digital, offline pulse shape discrimination and pulse timing method were used to measure five different samples of varying mass and burnup. The number of (neutron, neutron) correlations were selectively analyzed versus plutonium mass and a clear, increasing trend was observed. Additionally, the measurement scenarios were modeled using the MCNP-PoliMi code and good agreement was observed between the measured and simulated cross-correlation functions.

Measurement and simulation of neutron/gamma-ray cross-correlation functions from spontaneous fission

For the first time, a technique is presented for the measurement of total and separate neutron and gamma-ray cross-correlation functions from a spontaneous fission source. The cross-correlation functions are unique for given material-geometry configuration, thus represent signatures that can be used for the identification of radioactive materials. The measurement technique allows for the collection of fast coincidences within a time window of the order of a few tens of nanoseconds. A digital pulse shape discrimination technique is used, which allows for the accurate acquisition of the coincidences in all particle combinations. Specifically, separate neutron–neutron, neutron–gamma-ray, gamma-ray–neutron, and gamma-ray–gamma-ray coincidences are acquired with two liquid scintillation detectors. The measurements are compared to results obtained with the MCNP-PoliMi code, which simulates neutron and gamma-ray coincidences from a source on an event-by-event basis. This comparison leads to relatively good qualitative agreement. The measurements and simulations of the separate neutron and gamma-ray contributions to the total cross-correlations provide new signatures that can be obtained using existing experimental systems built to accurately identify nuclear materials. This research has direct applications in the areas of nuclear nonproliferation and homeland security.

Monte Carlo and analytical models of neutron detection with organic scintillation detectors

This paper presents a new technique for the analysis of neutron pulse height distributions generated in an organic scintillation detector. The methodology presented can be applied to techniques such as neutron spectrum unfolding, which have a variety of applications, including nuclear nonproliferation and homeland security. The technique is based on two independent approaches: (i) the use of the MCNP-PoliMi code to simulate neutron detection on an event-by-event basis with the Monte Carlo method and (ii) an analytical approach for neutron slowing down and detection processes. We show that the total neutron pulse height response measured by the organic scintillators is given by the sum of a large number of different neutron histories, each composed of a certain number of neutron scatterings on hydrogen and/or carbon. The relative contributions of each of these histories are described for a cylindrical liquid scintillator BC-501A. Simulations and measurements of neutron pulse height distributions are essential for neutron spectrum unfolding procedures.

Identification of shielded neutron sources with the liquid scintillator BC-501A using a digital pulse shape discrimination method

Fast and unambiguous identification of shielded neutron sources is of paramount importance in nuclear nonproliferation and international safeguards applications. Usually, the identification of neutron spectra relies on unfolding procedures that are unstable and time consuming. In this paper, we present a new application of an existing technique, which can be used for neutron source identification by direct analysis of the pulse height spectrum, i.e., without resorting to energy spectrum unfolding. This technique is based on the acquisition of neutron and γ-ray pulses by using a liquid scintillation detector and a fast waveform digitizer. An optimized digital pulse shape discrimination method based on standard charge integration is used to discriminate neutrons from γ-rays. The neutron pulses are then analyzed to generate a pulse height distribution. The accuracy of the neutron source identification technique was tested on three neutron sources: Cf-252, Am–Be, and Am–Li. Several source–shielding configurations were tested to assess the influence of potential shielding of the source and the sensitivity of the technique. For this investigation, lead and polyethylene shielding blocks were used. The measured pulse height distributions were compared with the distributions simulated with the MCNP-PoliMi code, and very good agreement was obtained. The results show that for both shielded and unshielded configurations the identification of Cf-252, Am–Be, and Am–Li shielded sources is easily achievable by direct analysis of the measured pulse height distributions, without the need of subsequent neutron energy spectrum unfolding.

Monte Carlo Modeling of Delayed Neutrons from Photofission

We describe the implementation of algorithms related to the delayed neutron production in photon-induced fission on actinides. The algorithms are based on data from experiments and have been implemented in the MCNP-PoliMi code. The modified code is being used to design and analyze methods to identify concealed highly enriched uranium with a system based on the use of photon interrogation and scintillation detectors.

Simulation of correlated counts from an Am–Li source

In this paper, we describe the implementation of a new source option in the Monte Carlo code MCNP-PoliMi. The code can be used to simulate time correlation measurements with neutrons and photons in plastic scintillation detectors. The new source describes the emission of neutrons and gamma-rays from an Am–Li (alpha, n) source. The simulations performed with the modified MCNP-PoliMi code are compared to corresponding measurements, and the reasons for disagreement are identified and discussed.

Calculation of gamma multiplicities in a multiplying sample for the assay of nuclear materials

The multiplicities, or factorial moments, of the distribution of the number of neutrons emerging from a fissile sample can be used to identify and quantify fissile isotopes, in particular even-N isotopes of transuranic elements. In fact, the spontaneously emitted source neutrons can induce further fissions in the sample, thereby changing the number distributions of the neutrons leaving the sample, and therefore their multiplicities. The multiplicities increase monotonically with sample mass, hence the measurement of the multiplicities can be used to quantify the sample mass. Analytical expressions for multiplicities that include induced fission effects have been derived for neutrons in the past. These expressions are given as functions of the probability of induced fission per neutron, and have been investigated both by Monte Carlo methods and in experiments using thermal neutron detectors. The object of this paper is to derive analytical formulae for the multiplicities of the gamma photons emitted by both spontaneous and induced fissions, and to perform a quantitative analysis. In addition, neutron and gamma multiplicities are calculated by Monte Carlo simulation using a modified version of the MCNP-PoliMi code. Good agreement is found between the analytical formulae and the Monte Carlo results. The results show the potential advantage of using gamma multiplicities when compared to neutron multiplicities: their higher quantitative values may, in principle, have the effect of leading to a larger sensitivity on the sample mass when compared to the analysis based on neutrons alone.

05/01296 Use of the MCNP-polimi code for time-correlation safeguards measurements

05/01296 Use of the MCNP-polimi code for time-correlation safeguards measurements

Monte Carlo analysis of neutron detection with a BaF/sub 2/ scintillation detector

This work presents the results of investigations aimed at simulating the response of a barium fluoride (BaF/sub 2/) detector to neutrons and photons. The simulations are performed with the MCNP-PoliMi code, a modification of MCNP-4C. The simulation results are compared to time-of-flight measurements performed with the nuclear materials identification system (NMIS). In particular, the neutron detection capabilities of the BaF/sub 2/ scintillator are analyzed.

MCNP-PoliMi: a Monte-Carlo code for correlation measurements

The Monte-Carlo simulation of correlation measurements that rely on the detection of fast neutrons and photons from fission requires that particle interactions in each history be described as closely as possible. The MCNP-PoliMi 1 1 PoliMi stands for Politecnico di Milano. code has been developed from the standard MCNP code to simulate each neutron–nucleus interaction as closely as possible. In particular, neutron interaction and photon production are made correlated and correct neutron and photon fission multiplicities have been implemented. The code output consists in relevant information about each collision, for example the type of collision, the collision target, the energy deposited, and the position of the interaction. A post-processing code has also been developed and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes. The application of the MCNP-PoliMi code to simulate neutron and gamma ray detection in a plastic scintillator is presented.

Use of the MCNP-polimi code for time-correlation safeguards measurements

Active nuclear safeguards measurements that rely on the time correlation between fast neutrons and gamma rays from the same fission are becoming a useful technique. In previous works we have shown the feasibility of this method, in conjunction with the use of the well-known MCNP simulation code and the use of artificial neural networks, to estimate the mass and enrichment of fissile samples enclosed in special, sealed containers.In a more recent works, we pointed out some features of MCNP that represent a drawback in the simulation of correlation measurements. In fact, MCNP is not intended for obtaining second order moments. Therefore, to achieve an agreement between the simulations and the experiment, we resorted to the use of effective physical parameters (for example detection threshold and light output). Recently, by suitably modifying MCNP, we developed the MCNP-PoliMi code, which attempts to simulate the physics of each interacton more realistically.In this paper, we present the results of the calibration of a plastic scintillator using a Cf-252 source, traditionally used in the above-mentioned nuclear safeguards experiments. We show that the physical parameters found by calibration and used in the MCNP-PoliMi simulations are reasonable for this type of detector, and that the correlation functions thereby obtained are in good agreement with the experiment.