Monoenergetic Neutron Sources Research Articles

Determination of neutron spectrum based on artificial neural network using liquid scintillation detector EJ-301.

This paper focuses on the neutron spectrum measurement using a liquid scintillation detector, where the neutron spectrum could be identified and unfolded from the light output distribution of the EJ-301 liquid scintillation detector through a linear artificial neural network (ANN). The response functions of the EJ-301 detector for monoenergetic neutron sources, as well as the light outputs, have been simulated and calculated by Monte Carlo procedure FLUKA. The linear ANN was trained and tested through the simulated data, where response functions were set as the input of ANN and the corresponding neutron spectra were output. Therefore, the neutron spectrum-unfolding model was created. This spectrum-unfolding model was tested through the light outputs induced by monoenergetic neutrons and the random superposition of them. Unfolding results show that this model could identify the information of the neutron spectrum accurately from the light outputs of a liquid scintillation detector. Moreover, the EJ-301 detector was used to measure the radioactivity of 252Cf, and the pulse height distribution induced by neutrons was derived through the charge-comparison method to remove the influence of gamma rays. The measured pulse height distribution was unfolded by the trained model, and measured results show that the unfolded neutron spectrum of 252Cf was consistent with the reference one. This paper presents the feasibility that the unknown neutron spectrum could be identified and confirmed through a linear neural network trained by simulated monoenergetic neutron response functions, which could be a candidate of choice for the determination of the neutron spectrum.

Measurement of the Response Function of a Custom Plastic Scintillator Using Monoenergetic Neutron and Proton Sources

Measurement of the Response Function of a Custom Plastic Scintillator Using Monoenergetic Neutron and Proton Sources

Optimization and experimental validation of deuterium-tritium neutron monoenergetics for diamond detector testing

Diamond detectors are excellent tools for measuring deuterium-tritium (D-T) neutron energy spectra. It has great potential for application in the field of magnetic confinement fusion. However, the current experimental values of the energy resolution of diamond detectors are much lower than the theoretical estimates. The purpose of this study is to find a method for experimental optimization. For this purpose, the D-T neutron energy distribution of the accelerator is theoretically analyzed and simulated. A method to increase the deuteron energy to obtain a better monoenergetic neutron source is proposed. This strategy was validated experimentally. The energy resolution of diamond for D-T neutrons achieves 1.21%, a Full Width at Half Maximum of 93 keV at 7.65 MeV. It is the highest resolution recorded to date. These results demonstrate the feasibility of the optimization method.

DEVELOPMENT AND TESTING OF A COMPUTER MODEL OF A SYSTEM FOR FORMING NEUTRON FLUXES ON A LINEAR ELECTRON ACCELERATOR

A computer model of the system for forming neutron fluxes at the output of a linear electron accelerator has been developed in the Geant4 programming environment. The system for forming neutron fluxes consists of a monoenergetic neutron source, a graphite reflector, a polyethylene moderator, lead protection and detector. With the help of the model, a number of virtual experiments on 107 primary neutrons were carried out for the energy range of the neutron source from 0.1 to 1.1 MeV. The dependence of the ratio of the number of neutrons falling on the detector with a reflector to the number of neutrons without a reflector is determined, on the energy of the neutron source. Energy spectra of neutrons falling on the detector are determined. It was established that the graphite reflector not only increases the number of particles falling on the detector, but also moderated the energy spectrum of neutrons. The neutron background in the accelerator bunker when using a reflector decreases from 2 to 7 times, depending on the energy of the neutron source.

Measurement of cross section for the 232Th(n, f) reaction using a time projection chamber

Accurate cross sections of neutron induced fission reactions are needed in the development of nuclear science and technology. The time projection chamber (TPC) is viewed as the potential detector to improve the measurement accuracy of the fission cross sections based on its track reconstruction and particle identification capabilities. The TPC developed by the back-streaming white neutron source (Back-n) team of China Spallation Neutron Source (CSNS) was introduced to verify its ability in measuring the cross section of fission reaction. Using our TPC, and based on the mono-energetic neutron source at Peking University and 232Th/238U samples, we have measured the cross section of the 232Th(n, f) reaction at 5.0 MeV. The fission events have been well distinguished from the background alpha events. The present cross section is consistent with the evaluation data from different libraries, which indicates that the fission cross section measurement using our TPC is feasible. By reducing the distance between the two fission samples, the large uncertainty of the present result will be reduced in future experiments.

Development of a new Monte-Carlo neutron transport code for radiation shielding

A continuous energy Monte-Carlo method-based neutron transport code for fixed source in 3D solid geometry has been developed. The main objective is to integrate this code with recently developed Monte-Carlo gamma shielding code MCBLD, so as to build an in-house code for coupled neutron-gamma shielding study. Also it is intended to use of the new ENDF data library (ENDF/B-VIII) in this code. This paper gives an overview of the code in terms of processing neutron interaction data and numerical scheme to track scattered neutrons. Present capability of the code is demonstrated by presenting fluence due to mono-energetic neutron source as well as widely used spontaneous neutron source from 252Cf isotope, placed at the centre of different size iron spheres. Results are also generated from general purpose open-source Monte-Carlo code FLUKA. Overall good agreement between the two validates the numerical implementation of angle and energy dependent neutron cross-section data in this code.

Neutron spectroscopy with TENIS using an artificial neural network

In this research, a ThErmal Neutron Imaging System (TENIS) consisting of two perpendicular sets of plastic scintillator arrays for boron neutron capture therapy (BNCT) application has been investigated in a completely different approach for neutron energy spectrum unfolding. TENIS provides a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions, but in the present study, the 70-pixel thermal neutron images have been used as input data for unfolding the energy spectrum of incident neutrons. Having generated the thermal neutron images for 109 incident mono-energetic neutrons, a 70 × 109 response matrix has been generated using the MCNPX2.6 code for feeding into the artificial neural network tools of MATLAB. The errors of the final results for mono-energetic neutron sources are less than 10% and the root mean square error (RMSE) for the unfolded neutron spectrum of 252Cf is about 0.01. The agreement of the unfolding results for mono-energetic and 252Cf neutron sources confirms the performance of the TENIS system as a neutron spectrometer.

COMBINED METHOD OF BULK MATERIAL SHIELDING EVALUATION FOR 200 MEV HIGH ENERGY NEUTRON SOURCE USING PHITS MODELLING AND PARTIAL DENSITY

Neutron encounters difficulties in shielding protection. Thus, many researchers have performed simulation and experimental research on neutron shielding materials. The characteristic of materials is highly dependent on neutron energy. The evaluation of neutron shielding for various materials, such as iron, concrete, aluminum, and borated polyethylene (BPE), was conducted in this paper through simulation using a Monte Carlo code of PHITS 3.27 and calculation via partial density method. A mono-energetic neutron source with an energy of 200 MeV is emitted perpendicular to the shielding material with a thickness of 105 cm. The parameters measured in this analysis include flux, fast neutron removal cross-section, neutron depth dose, ambient dose H*(10) equivalent, and neutron dose reduction factor (RF). Results show that iron is a good material against high-energy neutron and secondary photon radiation at the energy range with the highest removal cross-section and the lowest RF value (0.39), followed by concrete, BPE, and aluminum. The integrated fluence and effective dose profiles were consistent with previous results in the literature. Benchmarking calculation of neutron dose RF was conducted with other publications and was in good agreement within the value range.

A method to measure the quenching factor for recoil energy of oxygen in bismuth germanium oxide scintillators

Bismuth germanium oxide (Bi4Ge3O12, BGO) scintillation crystals are widely used as detectors in the fields of particle physics and astrophysics due to their high density, and thus higher efficiency for gamma-ray detection. Owing to their good chemical stability, they can be used in any environment. For rare-event searches, such as dark matter and coherent elastic neutrino-nucleus scattering, BGO crystals are essential to comprehend the response of nuclear recoil. In this study, we have analyzed the events of neutron elastic scattering with oxygen in BGO crystals. Then, we have measured the quenching factor for oxygen recoil energy in the BGO crystal as a function of recoil energy by using a monoenergetic neutron source.

Templates of expected measurement uncertainties for total neutron cross-section observables

This paper provides a template of expected uncertainties and correlations for measurements of total neutron cross-section observables by transmission. Measurements with time-of-flight and mono-energetic neutron sources are covered. The information required for evaluations in the resonance region and high energy region is detailed, along with the template of uncertainties and correlations that can be used in the absence of other information.

Templates of expected measurement uncertainties for (n, xn) cross sections

A template is provided for evaluating experimental uncertainties for neutron elastic and inelastic scattering cross sections andγ-ray production cross sections from (n, xn) measurements at laboratories with monoenergetic or white neutron sources. A typical range of uncertainties is presented for experiments detecting the scattered neutrons or the resulting de-excitationγrays based on a survey of available data and input from many experimentalists and theorists with extensive knowledge in the field. Models commonly used to evaluate the resulting cross-sections are also discussed. Suggestions are made regarding what experimental and uncertainty information is needed for data evaluations and should be included when reporting experimental (n, xn) cross sections. Uncertainty values and correlations are recommended if these values cannot be estimated for past data from the literature.

Excitation functions of neutron-induced threshold reactions in Au, Bi, Ta measured using 30–94 MeV quasi mono-energetic neutron sources

Activation measurements of neutron-induced threshold reactions were studied employing 7Li(p,n) quasi mono-energetic neutron sources in the neutron energy range of 30–94 MeV, where scarce experimental and evaluated data exist. A novel iterative neutron background subtraction approach is presented for the first time and its successful application is demonstrated on the analysis of the excitation functions of (n,xn) reactions in Au, Bi and Ta with x up to 10. Some of them are on the High Priority Nuclear Data Request List (HPRL). A comprehensive comparison of the measured data with theoretical predictions using the TALYS code is presented. While the calculated cross-sections agree with the measurements for lower-threshold reactions, severe discrepancies are observed for higher (n,xn) reactions for any of the TALYS model parameters.

IMPACT OF THE ENDF/B-VIII.0 LIBRARY ON MCNP6.2-COMPUTED AMBIENT AND PERSONAL DOSE EQUIVALENT CONVERSION COEFFICIENTS.

The 2018 release of the Evaluated Nuclear Data File (ENDF)/B-VIII.0 library initiated several examinations of the impact of the revised nuclear data on Monte Carlo models. This study used version 6.2 of the Monte Carlo N-Particle® code (MCNP6.2) with the ENDF/B-VI and ENDF/B-VIII.0 libraries to compute conversion coefficients for fluence-to-ambient and fluence-to-personal dose equivalent from neutron energy groups described by an International Atomic Energy Agency 53-bin structure and 47 monoenergetic neutron sources listed in International Commission on Radiological Protection (ICRP) 74. The MCNP6.2 models with ENDF/B-VI data were validated against results published in 2005 by Veinot and Hertel. Conversion coefficients computed with MCNP6.2 and ENDF/B-VIII.0 slightly underestimated the ICRP 74 values but were within ICRP-specified tolerances and do not justify revising the ICRP coefficients. Conversion coefficients for personal dose equivalent were computed with MCNP6.2 and ENDF/B-VIII.0 at four angles of incidence greater than those available in ICRP 74. These new coefficients are relevant to radiation protection studies of interest to US defence-focused organisations.

Evidence of fast neutron detection capability of the CLLB scintillation detector

For the first time we report the ability of the CLLB scintillator to detect and discriminate fast neutrons (2–10 MeV energy range) from gamma-rays and thermal neutrons, using a digital pulse shape discrimination technique. Firstly, we conducted measurements in single mode with an Am–Be source, unshielded and shielded, and we tagged a new cluster of events as the fast neutron-induced zone. The Figure of Merit obtained for the fast neutron/gamma discrimination was 1.1. Then, in order to better characterize these new cluster of events induced by fast neutrons, we conducted measurements in coincidence mode. We used a mono-energetic pulsed neutron source, at the Van de Graaff accelerator facility at the Legnaro National Laboratories (INFN-LNL, Legnaro-Italy), and also, a continuous neutron spectrum (using a 252Cf source). Combining all experimental results with a Monte Carlo model of the CLLB detector, built using the GEANT4 tool kit, it was possible to obtain the light output function of 7Li ions in the CLLB crystal (dL/dE = κ; κ = 0.36 ± 0.02). Moreover, the intrinsic fast neutron efficiency was modeled by Monte Carlo, obtaining similar values 1.5%–2% for neutrons between 2 MeV and 10 MeV. Finally, an application of this new feature, in the field of nuclear security and nuclear safeguards, is given.

TISSUE-EQUIVALENCE OF H2 GAS FOR MICRODOSIMETRY IN NEUTRON FIELDS: A GEANT4 MONTE CARLO STUDY.

Hydrogen (H2) as filling gas in Tissue-equivalent proportional counter (TEPC) for measurements of microdosimetric distributions in neutron fields is investigated using the Monte Carlo-based Geant4 toolkit. The neutron fields considered are monoenergetic neutrons (1keV-14MeV) and ISO reference neutron sources 241Am-Be, 241Am-B, 252Cf and 252Cf+D2O. Based on the energy deposited in the gas cavities (tissue-equivalent (TE) propane and H2) of the TEPC, density scaling correction, ${f}_{\rho }$, to be applied on H2 gas to achieve tissue-equivalence is calculated for the above sources at 1μm site size. The calculated value of ${f}_{\rho }$ for the ISO-neutron sources is 0.40, which suggests that energy deposition in the gas cavities is predominantly due to crossers. In the case of monoenergetic neutrons, depending upon the energy, ${f}_{\rho }$ is in the range of 0.11-0.45. The study shows that the TE propane and H2-based microdosimetric distributions are comparable when the density of H2 is scaled appropriately. Mean quality factor $\overline{Q}$ calculated based on microdosimetric distribution increases initially with neutron energy up to 100keV and thereafter decreases with energy. Depending upon the neutron energy, TE propane and H2-based $\overline{Q}$ values are comparable to within 3-15%, and for the ISO sources, the comparison is within 5-8%. For the ISO neutron sources, $\overline{Q}$ values are in the range of ~10-15.

Thin target neutron yield for Li(p,n) reaction with thick Ta target as stopping material

The neutron spectral yields from the Li(p,n) reaction has been considered as an important study due to its potential use as mono-energetic fast neutron source. This reaction has been studied at energies nearing Coulomb barrier for mono-energetic neutrons from 7Be ground state emissions and at energies beyond few tens of MeV to get the quasi mono-energetic fast neutrons, depending on their specific applications such as cross section studies, calibration of neutron detectors, etc. However, the number of measurements at proton energy of 8–20 MeV is sparse in the literature. In this energy range, the mono-energetic neutrons are strongly interfered by emissions from other excited states distorting the emission neutron spectra. Majority of the earlier measurements were carried out using carbon as the backing material during evaluation of neutron yields to minimize the interference of neutrons from the backing material. Here, a commonly used thick Ta target was explored as the backing material for the thin Li target and the neutron yields were measured with solid polymeric neutron track detector, CR-39. This is a passive technique and does not require any associated instrumentation during the irradiation process. This work presents the estimation of thin target neutron yields from Li target by subtracting the neutron yields contributed by the supporting material (Ta), which was separately measured in the same physical conditions.

Experimental evaluation of the Geant4-calculated response functions of a Bonner sphere spectrometer on monoenergetic neutron sources

The response function of a Bonner sphere spectrometer (BSS) plays a crucial role in deriving reliable energy distributions of neutron fluence from the experimental count of each sphere by using unfolding techniques. When using Monte Carlo codes to calculate the responses of a BSS, the factors of Bonner spheres descripted in the codes, e.g., material density and geometrical parameters, will influence the accuracy of the calculated responses. Therefore, experimental evaluation of the calculations using neutron sources is essential for establishing the response matrix. A Monte Carlo code Geant4 was used to calculate the responses of a BSS based on a 3He gas counter. The uncertainties of the calculated responses were analyzed after considering some factors, including the air gap between the gas counter and spherical moderators, the mass density of moderators, and the 3He gas pressure in the counter. Good agreement was found between the calculated responses (within uncertainties) and the experimental results on six ISO recommended monoenergetic neutron sources with energies from 144 keV to 14 MeV. To evaluate the capability of the Geant4 code in establishing the response matrix of a BSS, the calculated responses before and after experimental evaluation were illustrated and compared with the results obtained by using the Monte Carlo code MCNP5 for the BSS response calculation. These comparison results provide primary support of the effectiveness of the Geant4 code in establishing the response matrix of a BSS.

Development of the real-time double-ring fusion neutron time-of-flight spectrometer system at HL-2M

A real-time double-ring neutron time-of-flight (TOFII) spectrometer system has been proposed to achieve plasma diagnosis on HL-2M tokamak with a relatively high count rate and sufficient energy resolution. The TOFII system is in its development stage, and this work describes its characteristics in terms of design principle, system structure, electronic system design, preliminary tests, and neutron transport simulation. The preliminary test results illustrate that the TOFII system can demonstrate the real-time dynamic spectrum every 10 ms. The results also show that based on the support vector machine method, the n–γ discrimination algorithm achieves the discrimination accuracy of 99.1% with a figure of merit of 1.30, and the intrinsic timing resolution of the system is within 0.3%. The simulated flight time spectrums from 1 to 5 MeV are obtained through the Monte Carlo tool Geant4, which also provide the reasonable results. The TOFII system will then be calibrated on mono-energetic neutron sources for further verification.

Characterization Study of Fast Neutron Sources Based on Proton Accelerators at KOMAC

We introduce two neutron sources based on proton accelerators; a quasi monoenergetic neutron source driven by a 1.7 MV tandem accelerator and a pulsed white neutron source based on a 100 MeV linear proton accelerator, which are installed at Korea Multi-purpose Accelerator Complex (KOMAC) of Korea Atomic Energy Research Institute (KAERI). The neutron sources are characterized in terms of Monte Carlo simulations, and experiments based on various type of detectors and activation methods. In the manuscript, specifications of the neutron sources and the partial results of characterization study including energy and flux density measurements of the neutron sources are presented.

A Hierarchical Bayesian Approach to Neutron Spectrum Unfolding With Organic Scintillators

We propose a hierarchical Bayesian model and a state-of-the-art Monte Carlo sampling method to solve the unfolding problem, i.e., to estimate the spectrum of an unknown neutron source from the data detected by an organic scintillator. Inferring neutron spectra is important for several applications, including nonproliferation and nuclear security, as it allows the discrimination of fission sources in special nuclear material (SNM) from other types of neutron sources based on the differences of the emitted neutron spectra. Organic scintillators interact with neutrons mostly via elastic scattering on hydrogen nuclei and therefore partially retain neutron energy information. Consequently, the neutron spectrum can be derived through deconvolution of the measured light-output spectrum and the response functions of the scintillator to monoenergetic neutrons. The proposed approach is compared to three existing methods using the simulated data to enable controlled benchmarks. We consider three sets of detector responses. One set corresponds to a 2.5-MeV monoenergetic neutron source and two sets are associated with (energywise) continuous neutron sources ( <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">252</sup> Cf and <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">241</sup> AmBe). Our results show that the proposed method has similar or better unfolding performance compared with other iterative or Tikhonov regularization-based approaches in terms of accuracy and robustness against limited detection events while requiring less user supervision. The proposed method also provides a posteriori confidence measures, which offers additional information regarding the uncertainty of the measurements and the extracted information.