Neutron Pulse Height Distributions Research Articles

Simulation of the direct geometry spectrometer for neutron time of flight based on the Monte Carlo method to calculate the energy spectrum of the neutron source

Abstract In the present study, simulation of a direct geometry spectrometer for neutron Time-of-Flight (nTOF) is performed based on the Monte Carlo method to calculate the energy spectrum of the neutron source. To this end, the ability to simulate the simultaneous emission (such as the simultaneous emission of the neutron and gamma particles from the 241Am-9Be neutron source) and the neutron Pulse Height Distribution (nPHD) were added to the MCNPX computer code. In addition, a post-processing software was developed to analyze the massive amounts of the data in the output of the PTRAC card. The application of the nTOF as a direct method to calculate the neutron spectrum eliminates the error induced by the unfolding approach. Therefore, the direct methods for neutron spectroscopy such as nTOF give the results with high energy resolution and efficiency. The simulation of the nTOF measurement gives the opportunity to investigate the effect of the parameters such as the distance, angle, and detector type on the results. Therefore, it leads to select the best arrangement of the equipments to measure the neutron energy spectrum using the nTOF method. The results of the simulation in the present paper have an acceptable agreement with the standard ISO-8592 spectrum of the 241Am-9Be neutron source.

Neutron spectroscopy with soft computing: unfolding of the neutron energy spectrum using the developed computer code based on Adaptive Group of Ink Drop Spread (AGIDS)

This paper presents the developed computer codes based on the Adaptive Group of Ink Drop Spread (AGIDS) algorithm [1] for the reconstruction of the energy spectrum of the neutron source. The required data are generated via simulation of the neutron pulse height distributions due to randomly generated energy spectra with the MCNPX-ESUT computer code [2]. The simulated neutron pulse height distributions and corresponding randomly generated energy spectra are the input and output data of the developed computer code, respectively. As a case study, the 241Am-9Be neutron source is studied and the simulation of the neutron pulse height distribution of the NE-213 liquid organic scintillator is performed using the MCNPX-ESUT computer code. It is shown that the reconstructed energy spectrum of the 241Am-9Be neutron source has an excellent agreement with the reference spectrum provided by the International Standard Organization for neutron spectroscopy (ISO) [3].

Neutron spectroscopy with soft computing: development of a computational code based on Support Vector Machine (SVM) for reconstruction of neutron energy spectrum

This paper presents a developed computational code based on Support Vector Machine (SVM) for reconstruction of energy spectrum of neutron source. To reconstruct unknown energy spectrum using known neutron pulse height distribution, the developed machine is trained by known neutron pulse height distribution of detector and corresponding energy spectrum of neutron source. Validation and testing are the next steps to verify the validity of the calculations done with the developed computational code. The calculated neutron pulse height distributions due to randomly generated energy spectrum using MCNPX-ESUT (MCNPX-Energy engineering of Sharif University of Technology) computational code are used as input data of the developed computational code for reconstruction of energy spectrum of neutron source. The considered energy spectrums of neutron source are output data. In the present study, the simulation of neutron pulse height distribution of 241Am-9Be neutron source in the NE-213 liquid organic scintillator is performed using MCNPX-ESUT computational code. In the constructed matrix form of Fredholm integral equation, the effective response matrix is usually close to singular or badly scaled and its condition number is so high. The solution of inverse problem in which the matrix has high condition number does not give accurate results. The motivation of the present study is the development of computational coded based on soft computing algorithm for reconstruction of energy spectrum of the neutron source with high accuracy. To this end, the computational code based on Support Vector Machine (SVM) is developed, where there is no need to solve the inverse problems. The unfolded energy spectra of 241Am-9Be neutron sources using the developed computational code have an excellent agreement with the reference spectrum provided by ISO. A comparison between the results of present study and results of developed computational code based on Artificial Neural Network (ANN) and Adaptive Neuro-Fuzzy Inference System (ANFIS) is performed.

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi–Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers.

Characterization of Fissile Assemblies Using Low-Efficiency Detection Systems

We have investigated the possibility that the amount, chemical form, multiplication, and shape of the fissile material in an assembly can be passively assayed using scintillator detection systems by only measuring the fast neutron pulse height distribution and distribution of time intervals $\Delta \text{t}$ between fast neutrons. We have previously demonstrated that the alpha ratio can be obtained from the observed pulse height distribution for fast neutrons. In this paper, we report that when the distribution of time intervals is plotted as a function of log $\Delta t$ , the position of the correlated neutron peak is nearly independent of detector efficiency and determines the internal relaxation rate for fast neutrons. If this information is combined with knowledge of the alpha ratio, then the position of the minimum between the correlated and uncorrelated peaks can be used to rapidly estimate the mass, multiplication, and shape of fissile material. This method does not require a priori knowledge of either the efficiency for neutron detection or the alpha ratio. Although our method neglects three-neutron correlations, we have used previously obtained experimental data for metallic and oxide forms of Pu to demonstrate that our method yields good estimates for multiplications as large as 2, and that the only constraint on detector efficiency/observation time is that a peak in the interval time distribution due to correlated neutrons is visible.

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The 241Am-9Be and 252Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions.

Evaluation of neutron light output response functions in EJ-309 organic scintillators

An accurate model of the nonlinear detector response of organic scintillators to neutrons is required to correctly simulate fast neutron detection, as well as interpret measured pulse height data. Several empirical and semi-empirical models are available to fit measured scintillator light output data. In this work, EJ-309 light output data from neutrons depositing 1.15–5.15MeV on hydrogen were analyzed using empirical models as well as semi-empirical models based on the work of Birks and Voltz. Although all tested models fit the experimental light output data well in the measured range, the models were observed to diverge in low-energy extrapolation. The models were then tested by comparing a measurement and MCNPX-PoliMi simulation of an EJ-309 detector response to fast neutrons from a 252Cf spontaneous fission source. The agreement between the measured and simulated pulse height distributions varied significantly depending on the light output model used. The best agreement between simulated and measured neutron pulse height distributions was achieved by using the Birks model. The bin-by-bin agreement was better than 5% over the range 0.08–2.18MeVee, and better than 10% from 2.18 to 3.13MeVee. The integral count rate over the range 0.08–3.14MeVee differed by less than 1% in absolute units.

Neutron spectrum unfolding using artificial neural network and modified least square method

In the present paper, neutron spectrum is reconstructed using the Artificial Neural Network (ANN) and Modified Least Square (MLSQR) methods. The detector's response (pulse height distribution) as a required data for unfolding of energy spectrum is calculated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Unlike the usual methods that apply inversion procedures to unfold the energy spectrum from the Fredholm integral equation, the MLSQR method uses the direct procedure. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry of neutron sources, the neutron pulse height distribution is simulated/measured in the NE-213 detector. The response matrix is calculated using the MCNPX-ESUT computational code through the simulation of NE-213 detector's response to monoenergetic neutron sources. For known neutron pulse height distribution, the energy spectrum of the neutron source is unfolded using the MLSQR method. In the developed multilayer perception neural network for reconstruction of the energy spectrum of the neutron source, there is no need for formation of the response matrix. The multilayer perception neural network is developed based on logsig, tansig and purelin transfer functions. The developed artificial neural network consists of two hidden layers of type hyperbolic tangent sigmoid transfer function and a linear transfer function in the output layer. The motivation of applying the ANN method may be explained by the fact that no matrix inversion is needed for energy spectrum unfolding. The simulated neutron pulse height distributions in each light bin due to randomly generated neutron spectrum are considered as the input data of ANN. Also, the randomly generated energy spectra are considered as the output data of the ANN. Energy spectrum of the neutron source is identified with high accuracy using both MLSQR and ANN methods. The results obtained from MLSQR and ANN methods for 252Cf and 241Am-9Be source are validated against the ISO spectrum. The unfolded neutron energy spectra from both MLSQR and ANN methods show a good agreement with the actual spectrum of 252Cf and 241Am-9Be source.

Neutron light output function and resolution investigation of the deuterated organic liquid scintillator EJ-315

In this paper we investigate the light response to fast neutrons and estimate the pulse height resolution of a deuterated liquid organic scintillator, EJ-315, considering the detector's non-linear light response to gamma-rays. Initially, collision data and a neutron beam trigger are recorded in coincidence mode, and incident neutron energy is calculated with a time-of-flight technique. Fast neutrons are further discriminated from gamma-ray background based on the scintillation material decay patterns using a pulse shape discrimination algorithm. A light response matrix composed of multiple neutron energy and their corresponding light outputs is derived. The pulse height resolution property of the EJ-315 is characterized utilizing the derivatives of the pulse height distributions with corrections of the measurements setup uncertainties. Additionally, the EJ-315's pulse height resolution is also characterized by comparing the smoothed derivatives of quasi-monoenergetic neutron pulse height distributions, given by the Peierls-formula-based analytic model, to match the measurement data. Results show rather consistent 10–13% pulse resolution for mono-energetic neutrons with kinetic energy above 2 MeV. The resolution decreases slightly with an increase in neutron energy indicating the improved resolution performance of EJ-315 in the higher energy events.

Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation.Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing.In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304–309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137Cs source is also compared with the experimental data.

Neutron detector based on Particles of 6Li glass scintillator dispersed in organic lightguide matrix

Most 3He replacement neutron detector technologies today have overlapping neutron–gamma pulse-height distributions, which limits their usefulness and performance. Different techniques are used to mitigate this shortcoming, including Pulse Shape Discrimination (PSD) or threshold settings that suppress all gammas as well as much of the neutrons. As a result, count rates are limited and dead times are high when PSD is used, and the detection efficiency for neutron events is reduced due to the high threshold. This is a problem in most applications where the neutron–gamma separation of 3He detectors had been essential. This challenge is especially severe for neutron coincidence and multiplicity measurements that have numerous conflicting requirements such as high detection efficiency, short die-away time, short dead time, and high stability. 6Li-glass scintillators have excellent light output and a single peak distribution, but they are difficult to implement because of their gamma sensitivity. The idea of reducing the gamma sensitivity of 6Li-glass scintillators by embedding small glass particles in an organic light-guide medium was first presented by L.M. Bollinger in the early 60s but, to the best of our knowledge, has never been reduced to practice. We present a proof of principle detector design and experimental data that develop this concept to a large-area neutron detector. This is achieved by using a multi-component optical medium (6Li glass particles attached to a glass supporting structure and a mineral oil light guide) which matches the indices of refraction and minimizes the absorption of the 395nm scintillator light. The detector design comprises a 10in. long tube with dual end readout with about 3% volume density of 6Li glass particles installed. The presented experimental data with various neutron and gamma sources show the desired wide gap between the neutron and gamma pulse height distributions, resulting in a true plateau in the counting characteristics similar to that of 3He detectors.

Plutonium metal vs. oxide determination with the pulse-shape-discrimination-capable plastic scintillator EJ-299-33

Neutron measurements can be used to distinguish plutonium in metal or oxide form, a capability that is of great interest in nuclear nonproliferation, treaty verification, and other applications. This paper describes measurements performed on well-characterized samples of plutonium oxide and plutonium metal using the pulse-shape-discrimination-capable plastic scintillator EJ-299-33. Results are compared to those obtained with a same-sized detector cell using the liquid scintillator EJ-309. The same optimized, digital pulse shape discrimination technique is applied to both detectors and the neutron pulse height distributions are compared. Results show that the EJ-299-33 plastics can be successfully used for plutonium measurements, where the gamma ray to neutron detection ratio is much higher than for typical radioactive sources. Results also show that EJ-299-33 detectors can be used to characterize plutonium samples, specifically to discriminate between plutonium metal and oxide.

Tomographic analysis of neutron and gamma pulse shape distributions from liquid scintillation detectors at Joint European Torus

The Joint European Torus (JET, Culham, UK) is the largest tokamak in the world devoted to nuclear fusion experiments of magnetic confined Deuterium (D)/Deuterium-Tritium (DT) plasmas. Neutrons produced in these plasmas are measured using various types of neutron detectors and spectrometers. Two of these instruments on JET make use of organic liquid scintillator detectors. The neutron emission profile monitor implements 19 liquid scintillation counters to detect the 2.45 MeV neutron emission from D plasmas. A new compact neutron spectrometer is operational at JET since 2010 to measure the neutron energy spectra from both D and DT plasmas. Liquid scintillation detectors are sensitive to both neutron and gamma radiation but give light responses of different decay time such that pulse shape discrimination techniques can be applied to identify the neutron contribution of interest from the data. The most common technique consists of integrating the radiation pulse shapes within different ranges of their rising and/or trailing edges. In this article, a step forward in this type of analysis is presented. The method applies a tomographic analysis of the 3-dimensional neutron and gamma pulse shape and pulse height distribution data obtained from liquid scintillation detectors such that n/γ discrimination can be improved to lower energies and additional information can be gained on neutron contributions to the gamma events and vice versa.

Validation of Geant4 and MCNPX-PoliMi simulations of fast neutron detection with the EJ-309 liquid scintillator

In organic liquid scintillators, the energy distribution of neutrons is indirectly measured by characterizing the pulse height distribution of light output produced through nuclear reactions within the detectors. However, the complexity of the light generation on these detectors makes simulating their response difficult with standard Monte Carlo techniques. This paper presents the response of 7.62-cm-long×7.62-cm-diameter and 12.7-cm-long×12.7-cm-diameter EJ-309 liquid scintillators, simulated using the Geant4 and MCNPX-PoliMi Monte Carlo codes. The simulations are restricted to the neutron transport and interaction models. Monoenergetic neutrons as well as fission neutrons emitted by a 252Cf spontaneous fission source are used in these simulations. The energy deposition of neutrons in organic liquid scintillators is dependent upon their collisions with hydrogen and carbon nuclei; therefore simulating neutron collisions on an event-by-event basis is important to generate neutron pulse height distributions. There are no default physics models in Geant4 hence proper selection of physics models is required for accurate simulation. In this paper, we compare Geant4 (ver. 4.9.5) and MCNPX-PoliMi (ver. 2.0) neutron pulse height distributions. The simulations are validated with time-of-flight (TOF) measurements obtained with 12.7-cm-long×12.7-cm-diameter EJ-309 detectors. Results show that both Geant4- and MCNPX-PoliMi-generated distributions are in good agreement with the measured data.

Simulation of Neutron Pulse Height Distributions with a Response Matrix Method

A response matrix method (RMM) is applied to Monte Carlo simulations to efficiently compute neutron pulse height distributions (PHDs) in organic scintillation detectors. The PHD calculations and their associated uncertainty are compared for a polyethylene-shielded and lead-shielded 252Cf source for three different techniques: fully analog MCNPX-PoliMi, the RMM, and the RMM with source biasing. The RMM with source biasing reduces computation time or improves the figure of merit on average by a factor of 600 for polyethylene shielding and a factor of 300 for lead shielding (when compared to the fully analog calculation). The simulated neutron PHDs show good agreement with the laboratory measurements, thereby validating the RMM.

D-D neutron-scatter measurements for a novel explosives-detection technique

A series of measurements has been completed that provides a benchmark for Monte Carlo simulations related to an algorithm for explosives detection using active neutron interrogation. The original simulations used in algorithm development, based on land-sea cargo container screening, have been adapted to model active neutron interrogation of smaller targets. These smaller-scale measurements are easily accomplished in a laboratory environment. Benchmarking measurements were completed using a D-D neutron generator, two neutron detectors, as well as a variety of scatter media including the explosives surrogate melamine (C3H6N6). Measurements included 90°, 120°, or 150° neutron scatter geometries and variations in source-detector shielding, target presence, and target identity. Comparisons of measured and simulated neutron fluxes were similar, with correlation coefficients greater than 0.7. The simulated detector responses also matched very closely with the measured photon and neutron pulse height distributions, with correlation coefficients exceeding 0.9. The experiments and simulations also provided insight into potential application of the new method to the problem of explosives detection in small objects such as luggage and small packages.

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.

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.

Design and performance of an array of plastic scintillators for detecting neutrons of 1.0 to 450 MeV

The design and performance of an array of plastic scintillator neutron counters are discussed. The array uses counters of three sizes to detect neutrons of 1.0 to 450 MeV, measuring their energy by time-of-flight and their production angles by counter identification and end-to-end timing. Information from 180 000 neutron-proton charge exchange events is used to investigate the resolution stability and internal consistency of the system. Neutron pulse height distributions for narrow bins of neutron energy between 1.25 and 313 MeV are presented and compared with the predictions of a Monte Carlo program.

Reactor Handbook, volume I, materials

In the present paper, neutron spectrum is reconstructed using the Artificial Neural Network (ANN) and Modified Least Square (MLSQR) methods. The detector's response (pulse height distribution) as a required data for unfolding of energy spectrum is calculated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Unlike the usual methods that apply inversion procedures to unfold the energy spectrum from the Fredholm integral equation, the MLSQR method uses the direct procedure. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry of neutron sources, the neutron pulse height distribution is simulated/measured in the NE-213 detector. The response matrix is calculated using the MCNPX-ESUT computational code through the simulation of NE-213 detector's response to monoenergetic neutron sources. For known neutron pulse height distribution, the energy spectrum of the neutron source is unfolded using the MLSQR method. In the developed multilayer perception neural network for reconstruction of the energy spectrum of the neutron source, there is no need for formation of the response matrix. The multilayer perception neural network is developed based on logsig, tansig and purelin transfer functions. The developed artificial neural network consists of two hidden layers of type hyperbolic tangent sigmoid transfer function and a linear transfer function in the output layer. The motivation of applying the ANN method may be explained by the fact that no matrix inversion is needed for energy spectrum unfolding. The simulated neutron pulse height distributions in each light bin due to randomly generated neutron spectrum are considered as the input data of ANN. Also, the randomly generated energy spectra are considered as the output data of the ANN. Energy spectrum of the neutron source is identified with high accuracy using both MLSQR and ANN methods. The results obtained from MLSQR and ANN methods for 252Cf and 241Am-9Be source are validated against the ISO spectrum. The unfolded neutron energy spectra from both MLSQR and ANN methods show a good agreement with the actual spectrum of 252Cf and 241Am-9Be source.