Monoenergetic Neutron Sources Research Articles

Estimating radiation protection factor (RPF) values for a simple surrogate vehicle using the MCNP6.1 code

Computational and experimental estimates of radiation protection factor (RPF) values are of significant interest to various defense-related organizations. Values of the operational quantity of ambient dose equivalent were computed, using version 6.1 of the Monte Carlo N-Particle® code (MCNP6.1), for 25 photon energies and 53 neutrons energies. Energy-dependent values for absorbed dose at a 10-mm depth in the 30-cm-diameter ICRU tissue-equivalent sphere, D*(10), were estimated using the kerma approximation for photon (MCNP F6:p tally) and neutron (MCNP F6:n tally) radiation fields, and then transformed to fluence-to-ambient dose equivalent conversion coefficients, H*(10)/ϕ, using appropriate quality factors, Q, and radiation fluences. MCNP6.1-computed values of H*(10)/ϕ were compared with tabulated values from ICRP Publication 74. The published conversion coefficients for photon radiation fields were within the error bounds of all MCNP6.1-calculated values. For neutron fields, the computed and published conversion coefficients agreed within calculated error bounds for neutron energies of 100 keV and above, while responses to neutron fields below 100 keV exhibited the same shape and were consistent with the kerma approximation applied in the computational model. The MCNP6.1 models were then modified to simulate the same radiation fields, but H*(10)/ϕ conversion coefficients were calculated in the shielded environment of a surrogate vehicle (steel cube)—in keeping with historical and recent studies—and the ratios of shielded and unshielded vehicle response functions were expressed as energy-dependent RPF values. The MCNP6.1-computed neutron RPF values were compared with two prior experimental studies involving a mono-energetic neutron source and a high-yield, short-duration fission neutron spectrum and found to agree within the uncertainty of the experimentally-measured values. Limitations of the computed RPF values are discussed.

Measurements of cosmic ray induced background neutrons near the ground using a Bonner sphere spectrometer

This report documents measurements of cosmic ray induced background neutrons using a new Bonner sphere spectrometer (BSS) based on a 3 He thermal neutron counter (SP9 type). Three spheres with a lead or copper layer were added to a BSS to enhance its sensitivity in the high-energy range above 20 MeV and enable the spectrometer to cover the entire energy range from thermal to multi-GeV. The response function of the spectrometer was calculated using the Monte Carlo code MCNP6. Below 20 MeV, the responses of polyethylene spheres were experimentally verified with 6 mono-energetic neutron sources . A neutron spectrum was measured near the ground at one location in Beijing. Also deduced from the spectrum were the rates of neutron fluence , ambient dose equivalent, and effective dose. • A Bonner sphere spectrometer with three extended spheres was employed to measure cosmic ray induced background neutrons at Beijing. • The MCNP6 code was used for the response function calculation. • Each count rate of the BSS was obtained after neutron flux variation correction. • A neutron spectrum was unfolded, and the deduced flux and dose rate values were also compared with other works.

The performance simulation of the LiH-SiC-based Fast Neutron Detector for harsh environment monitoring using Geant4 and TCAD

In this work, Lithium Hydride (LiH) and Silicon Carbide (SiC) based sensor for fast neutron detection has been reported for the first time. LiH is used as a hydrogenous converter material for converting neutrons to charged particles (recoil protons) and wide band-gap semiconductor material SiC is employed as a charged particle sensing detector in the form of a Schottky Barrier Diode (SBD). Geant4 Monte-Carlo simulations have been performed for the optimization of the LiH converter layer thickness for different mono-energetic neutron sources as well as Am-Be standard neutron source. The optimization of LiH thickness is essential to attain the maximum neutron detection efficiency. The simulation study has revealed that the maximum neutron detection efficiency of ∼ 0.1% can be achieved at ∼800μm thickness of LiH for the Am-Be neutron source. Furthermore, a device simulation software (Silvaco TCAD) has been utilized to investigate the radiation-induced effects on the electrical characteristics of a SiC-based SBD detector. The gamma and neutron irradiation induced damage model has been developed from the experimental deep level defects information reported in the literature for the TCAD simulation study. The effects on different device parameters viz., ideality factor, Schottky barrier height, series resistance, built-in-potential, effective carrier concentration, image force lowering, carrier removal rate, etc., due to 60Co-gamma (dose=100 Mrad) and 1 MeV equivalent neutron irradiation (fluence up to 1016 neutrons/cm2) have also been reported. The 1 MeV equivalent neutron fluence is the fluence of 1 MeV neutrons producing the same damage in a detector material as induced by an arbitrary particle fluence with specific energy distribution. This model will be helpful in predicting the performance and behavior of the SiC-based devices which are subjected to heavy irradiation. The study reveals that the effect on the electrical characteristics of SiC-based SBD detectors is tolerable up to the fluence of 1015 neutrons/cm2 at room temperature, but preventive measures must be adopted to reduce the leakage current at elevated temperatures.

Experimental characterization of a long counter for neutron fluence measurement

A De Pangher-type long counter was constructed at Peking University for neutron fluence measurement. The responses and effective centers of the long counter were calculated using the Monte Carlo method. The variations with neutron energy of the position of the effective center, calculated by the Monte Carlo code, was experimentally validated with a 241AmBe neutron source and several mono-energetic neutron sources ranging from 100 keV to 6 MeV. Long counter calibration was performed using a radionuclide source from Peking University, and neutron fluence measured by the calibrated long counter was successfully compared to values determined with reference instruments (two recoil proton counters and a 238U fission chamber). The relative deviations were lower than 6% along the whole energy range. The calibrated long counter was thus successfully applied in the calibration of a Bonner sphere spectrometer.

Design of a shielding collimator device for a small-angle monoenergetic neutron source

To obtain a small-angle monoenergetic neutron source, a shielding collimator device is designed for the neutron source generated by a neutron tube. The device is divided into the collimator and the capture cave. The collimator is made of three layers of stainless steel and borated polyethylene and is used to constrain neutrons in a small angle. The capture cave is used to increase the number of times neutron inelastic scattering occurs in the opposite direction of the radiation field, thereby reducing the proportion of scattered neutrons in the radiation field. Material thickness, aperture size, and the optimum structure of the capture cave were simulated using MCNP. The design features a neutron emission angle within a range of 3° and neutron fluxes in the radiation field, which are higher by two orders of magnitude than those outside the radiation field. This research has practical value for the generation of monoenergetic small-angle neutron sources and neutron applications.

Activation Measurements of Cross Sections for Ground and Isomeric States Production in Neutron Threshold Reactions on Y and Au

The paper describes cross-section measurements consisting of neutron activation using a 7Li(p,n) quasi mono-energetic neutron source and gamma-ray spectrometry performed at the Nuclear Physics Institute of the Czech Academy of Sciences in Řež, Czech Republic. The analysis of the excitation functions of neutron-induced threshold reactions for the energy range from 17 up to 34 MeV is presented. The cross sections for the population of both ground and isomeric states were measured in yttrium and gold. A comprehensive comparison of the measured cross sections with earlier experimental data from the EXFOR database, TALYS calculations, and evaluated nuclear data libraries is presented.

Use of SRIM and Garfield with Geant4 for the characterization of a hybrid 10B/3He neutron detector

This paper describes a method for more complete neutron detector characterization using Geant4’s Monte Carlo methods for characterizing overall detector response rate and Garfield interfaced with SRIM for the simulation of the detector’s raw pulses, as applied to a hybrid10B/3He detector. The Geant4 models characterizing the detector’s interaction with a 252Cf point source and parallel beams of mono-energetic neutrons (assuming ISO 8529 reference energy values) compare and agree well with calibrated252Cf measurements to within 6.4%. Validated Geant4 model outputs serve as input to Garfield+SRIM calculations to provide meaningful pulse height spectra. Modifications to Garfield for this work were necessary to account for simultaneous tracking of electrons resulting from proton and triton reaction products from a single3He neutron capture event, and it was further necessary to interface Garfield with the energy loss, range, and straggling calculations provided by SRIM. Individual raw pulses generated by Garfield+SRIM are also observed to agree well with experimentally measured raw pulses from the detector.

Investigation of enhancement in planar fast neutron detector efficiency with stacked structure using Geant4

Geant4 based Monte Carlo study has been carried out to assess the improvement in efficiency of the planar structure of Silicon Carbide (SiC)-based semiconductor fast neutron detector with the stacked structure. A proton recoil detector was simulated, which consists of hydrogenous converter, i.e., high-density polyethylene (HDPE) for generating recoil protons by means of neutron elastic scattering (n, p) reaction and semiconductor material SiC, for generating a detectable electrical signal upon transport of recoil protons through it. SiC is considered in order to overcome the various factors associated with conventional Si-based devices such as operability in a harsh radiation environment, as often encountered in nuclear facilities. Converter layer thickness is optimized by considering 109 neutron events of different monoenergetic neutron sources as well as 241Am-Be neutron spectrum. It is found that the optimized thickness for neutron energy range of 1–10 MeV is ~ 400 μm. However, the efficiency of fast neutron detection is estimated to be only 0.112%, which is considered very low for meaningful and reliable detection of neutrons. To overcome this problem, a stacked juxtaposition of converter layer between SiC layers has been analyzed in order to achieve high efficiency. It is noted that a tenfold efficiency improvement has been obtained—1.04% for 10 layers stacked configuration vis-a-vis 0.112% of single converter layer detector. Further simulation of the stacked detector with respect to variable converter thickness has been performed to achieve the efficiency as high as ~ 3.85% with up to 50 stacks.

Study of the 15N(p,n)15O reaction as a monoenergetic neutron source for the measurement of differential scattering cross sections

The 15N(p,n) reaction is a promising candidate for the production of monoenergetic neutrons with energies of up to 5.7 MeV at the facilities where the T(p,n)3He reaction cannot be used. The characteristic properties of this reaction were studied focusing on its suitability as a source of monoenergetic neutrons for the measurement of differential scattering cross sections in the neutron energy range of 2 MeV to 5 MeV . For this purpose differential and integral cross sections were measured and the choice of optimum target conditions was investigated. The reaction has already been used successfully to measure of elastic and inelastic neutron scattering cross sections for natPb in the energy range from 2 MeV to 4 MeV and for 209Bi and 181Ta at 4 MeV .

10B+ZnS(Ag) as an alternative to 3He-based detectors for Radiation Portal Monitors

Typical radiation portal monitor systems, RPM, deployed to detect illicit trafficking of radioactive materials include a set of gamma-ray detectors and neutron detectors. Usually the employed neutron detectors are pressurized 3 He-based neutron detectors tubes. Due the shortage of 3 He reported since 2009, the amount of 3 He available for use in gas proportional counter neutron detectors has become limited, while the demand has significantly increased, especially for homeland security applications. For this reason, many different alternatives are being investigated for its use in RPM systems. The aim of this work is to study a scintillation detector ZnS(Ag) mixed with highly enriched 10 B, 10 B+ZnS(Ag). Using Monte Carlo methods, MCNPX code, the response of two neutron detectors based on 10 B+ZnS(Ag), manufactured by BridgePort Instruments LLC with different geometries, were estimated by calculating the number of 10 B(n,α)7 Li reactions for 29 monoenergetic neutron sources. Measurements and models were made, and both detectors were compared. The importance of the distance with respect to the ground was studied. The response with a 252 Cf moderated neutron source (0.5 cm lead and 2.5 cm polyethylene) was calculated in order to compare with other studied alternatives in the USA by Pacific National Northwest Laboratory, PNNL. With these results we conclude that neutron detectors using 10 B+ZnS(Ag) are an interesting alternative for replacing 3 He detectors. From the analysis with MCNPX we propose an improvement in the detector design.

Neutron spectroscopy by thermalization light yield measurement in a composite heterogeneous scintillator

An exothermic neutron capture reaction can be used to uniquely identify neutrons in particle detectors. With the use of a capture-gated coincidence technique, the sequence of scatter events that lead to neutron thermalization prior to the neutron capture can also be used to measure neutron energy. We report on the measurement of thermalization light yield via a time-of-flight technique in a polyvinyl toluene-based scintillator EJ-290 within a heterogeneous composite detector that also includes 6Li-doped glass scintillator. The thermalization light output exhibits a strong correlation with neutron energy because of the preference for near-complete energy deposition prior to the 6Li(n,t)4He neutron capture reaction. The nonproportionality of the light yield from nuclear recoils contributes to the observed broadening of the distribution of thermalization light output. The nonproportional dependence of the scintillation light output in the EJ-290 scintillator as a function of proton recoil energy has been characterized in the range of 0.3–14.1MeV via the Birks parametrization through a combination of time-of-flight measurement and previously conducted measurements with monoenergetic neutron sources.

The design of “multiple energies from one source” in D-T neutron tube

To obtain multiple monoenergetic neutron sources and realize the on-site calibration of radiation monitoring equipment for nuclear-involved places, the structural characteristics and neutron source features of D-T neutron tube were analyzed; Monte Carlo method was adopted to simulate the effect of interaction between typical materials and different energy neutrons; multilayered shielding materials were combined and optimized to acquire the optimal scheme to shield the neutron sources from the neutron tube. On the base, a tapered alignment filtration construction was designed and Monte Carlo method was employed to simulate the effect of alignment construction. The result showed that the tapered alignment filtration construction can create monoenergetic neutrons including 14.1 MeV, 0.18 MeV and thermal neutrons and demonstrated good monochrome performance which provides multiple monoenergetic sources for the on-site calibration.

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.

Experimental Concept for a Precision Measurement of Nuclear Recoil Ionization Yields for Low Mass WIMP Searches

Understanding the response of dark matter detectors at the lowest recoil energies is important for correctly interpreting data from current experiments or predicting the sensitivity of future experiments to low mass weakly interacting massive particles. In particular, the ionization yield is essential for determining the correct recoil energy of candidate nuclear recoil events; however, few measurements in cryogenic crystals exist below 1 keV. Using the voltage-assisted calorimetric ionization detection technique with a mono-energetic neutron source, we show that it is possible to determine the ionization yield in cryogenic crystals down to an energy to 100 eV. This measurement will also determine the statistics of ionization production at these low energies.

The Breakup Cross Section of the D+D Reaction at 6.94 MeV

The D+D reactions are well known and widely used for a variety of purposes, mainly because of the use of the D(d , n )3 He reaction as a mono-energetic neutron source. The least studied of the D+D reactions is the D(d , n )pD reaction known as the deuteron breakup reaction, which produces a continuum of neutrons at energies below the monoenergetic peak. The neutron energy distribution as a function of angle for the cross section, , of the D(d ,n )pD reaction has been measured using a 6.94-MeV pulsed deuteron beam incident upon a D2 gas target. The time-of-flight technique was used to determine the energy of the neutrons detected in an array of two lithium glass scintillators and one NE-213 scintillator. The breakup cross section was determined as low as 225-keV neutron energy in the lithium glass detectors.

Energy Dependence of Fission Product Yields from 235U, 238U and 239Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varying degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for 235U, 238U and 239Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual-fission chamber and gamma-ray counted using shielded HPGe detectors for a period of 1-2 months to determine the yield of various fission products. To the extent possible all irradiation and counting procedures were kept the same to minimize sources of systematic errors. FPY have been determined at incident neutron energies of 0.6, 1.4, 2.4, 3.5, 4.6, 5.5, 8.9 and 14.8 MeV.

Maximum Likelihood Estimation of Neutron Source Location

Algorithms for locating the neutron source by neutron time-of-flight (TOF) measurement are established and discussed for monoenergetic and multienergetic neutrons in this paper. For the monoenergetic neutron source, the location of the source could be estimated by locating the position where the variance between the actual TOF and the calculation gains its minimum. For multienergetic neutrons the maximum likelihood (ML) method has been applied to process the time-correlation measurement. The efficiencies of location estimations are studies with simulations. In the simulations, the TOFs are acquired by time-correlation measurement of three neutron detectors surrounding the suspected area of the neutron source. The results indicate that the location of monoenergetic neutron sources could be estimated by the neutron TOF acquired by three detectors, while for multienergetic neutron sources, the likelihood method could be used to locate the most probable location, as well as its possible distribution of location. As the result of large-scale simulation and comparison, the ML estimation method is more effective than traditional methods, especially in conditions of low count rates or low resolution.

SU‐E‐T‐21: A D‐D Based Neutron Generator System for Boron Neutron Capture Therapy: A Feasibility Study

Purpose:To investigate the feasibility of a deuterium‐deuterium (DD) neutron generator for application in boron neutron capture therapy (BNCT) of brain cancerMethods:MCNP simulations were performed using a head phantom and a monoenergetic neutron source, which resembles the point source in a DD generator that emits 2.45‐MeV neutrons. Source energies ranging from 5eV to 2.45MeV were simulated to determine the optimal treatment energy. The phantom consisted of soft tissue, brain tissue, skull, skin layer, and a brain tumor of 5 cm in diameter. Tumor depth was varied from 5–10 cm. Boron‐10 concentrations of 10 ppm, 15 ppm, and 30 ppm were used in the soft/brain tissues, skin, and tumor, respectively. The neutron flux required to deliver 60 Gy to the tumor as well as the normal tissue doses were determined.Results:Beam energies between 5eV and 10keV obtained doses with the highest dose ratios (3.3–25.9) between the tumor and the brain at various depths. The dose ratio with 2.45‐MeV neutrons ranged from 0.8–6.6. To achieve the desired tumor dose in 40 minutes, the required neutron flux for a DD generator was between 8.8E10 and 5.2E11 n/s and the resulting brain dose was between 2.3 and 18 Gy, depending on the tumor depth. The skin and soft tissue doses were within acceptable tolerances. The boron‐neutron interaction accounted for 54–58% of the total dose.Conclusion:This study shows that the DD neutron generator can be a feasible neutron source for BNCT. The required neutron flux for treatment is achievable with the current DD neutron technology. With a well‐designed beam shaping assembly and treatment geometry, the neutron flux can be further improved and a 60‐Gy prescription can be accurately delivered to the target while maintaining tolerable normal tissue doses. Further experimental studies will be developed and conducted to validate the simulation results.

Neutron crosstalk between liquid scintillators

A method is proposed to quantify the fractions of neutrons scattering between liquid scintillators. Using a spontaneous fission source, this method can be utilized to quickly characterize an array of liquid scintillators in terms of crosstalk. The point model theory due to Feynman is corrected to account for these multiple scatterings. Using spectral information measured by the liquid scintillators, fractions of multiple scattering can be estimated, and mass reconstruction of fissile materials under investigation can be improved. Monte Carlo simulations of mono-energetic neutron sources were performed to estimate neutron crosstalk. A californium source in an array of liquid scintillators was modeled to illustrate the improvement of the mass reconstruction.

A method for using neutron elastic scatter to create a variable energy neutron beam from a nearly monoenergetic neutron source

Abstract This work describes preliminary investigation into the design of a compact, portable, variable energy neutron source. The proposed method uses elastic neutron scatter at specific angles to reduce the energy of deuterium–deuterium or deuterium–tritium (D–T) neutrons. The research focuses on D–T Monte Carlo simulations, both in idealized and more realistic scenarios. Systematic uncertainty of the method is also analyzed. The research showed promise, but highlighted the need for discrimination of multiply-scattered neutrons, either through a pulsed generator or associated particle imaging.