Monoenergetic Neutron Sources Research Articles

Detecting energy dependent neutron capture distributions in a liquid scintillator

A novel technique is being developed to estimate the effective dose of a neutron field based on the distribution of neutron captures in a scintillator. Using Monte Carlo techniques, a number of monoenergetic neutron source energies and locations were modelled and their neutron capture response was recorded. Using back propagation Artificial Neural Networks (ANN) the energy and incident direction of the neutron field was predicted from the distribution of neutron captures within a 6Li-loaded liquid scintillator. Using this proposed technique, the effective dose of 252Cf, 241AmBe and 241AmLi neutron fields was estimated to within 30% for four perpendicular angles in the horizontal plane. Initial theoretical investigations show that this technique holds some promise for real-time estimation of the effective dose of a neutron field.

Analysis of neutron scattering components inside a room with concrete walls

This paper describes the scattering corrections needed when neutron detectors are calibrated with a neutron point source at the center of a calibration room. The independence of scattering value from the geometric shape of the room was studied, which for more confidence in this case, two sets of rooms with same inner surface area were evaluated. The parameters that relate the air scattering part (A) and room-return part (R) with the additional contribution from scattered neutrons have been calculated for neutrons whose energy goes from 10−8 to 20MeV. These parameters were calculated using Monte Carlo method for 150, 200, 300, 407, 500, 800 and 1000cm-radius spherical cavity containing air. In the calculations, monoenergetic neutron sources were placed at the center of cavity, and then neutron fluences were determined at several distances of source to detector along the spherical cavity radius. The parameter R has been fitted as a function of surface area (or radius) of spherical room, and the related coefficient has been calculated for 16 monoenergetic neutron sources. It may produce a reasonable estimate (with difference <10%) for the contribution of walls scattering in any geometry of a calibration room.

Cross-section studies of important neutron and relativistic deuteron reactions

The cross-sections of relativistic deuteron reactions on natural copper were studied by the means of activation method. The deuteron beams produced by JINR Nuclotron (Russia) with energies from 1 GeV up to 8 GeV were used. Lack of such cross-sections prevents the usage of copper foils for beam integral monitoring. The copper monitors will help us to improve the beam integral determination during ADS studies. The yttrium samples are very suitable activation detectors for monitoring of neutron fields not only in the ADS studies. But experimental cross-section data for higher energy threshold neutron reactions are still missing. This situation is the reason why we have started to study neutron reactions on yttrium by the means of quasi mono-energetic neutron source based on NPI Řež cyclotron (Czech Republic).

Blow-off momentum from melt and vapor in nuclear deflection scenarios

For Earth-impacting objects that are large in size or have short warning times nuclear explosives are an effective threat mitigation response. Nuclear-based deflection works by means of conservation of momentum: as material is heated by incoming photons and neutrons it is ejected from the body which imparts momentum to the remaining mass of the asteroid. Predicting the complete response of a particular object is difficult, since the ejecta size and velocity distributions rely heavily on the unknown, complicated internal structure of the body. However, lower bounds on the blow-off momentum can be estimated using the melted and vaporized surface material. In this paper, we model the response of a one-dimensional SiO2 surface to monoenergetic soft X-ray, hard X-ray and neutron sources using Arbitrary Lagrangian–Eulerian radiation/hydrodynamic simulations. Errors in the blow-off momentum due to our hydrodynamic mesh resolution are quantified and inform zone sizing that balances numerical discretization error with computational efficiency. We explore deposited energy densities ranging from 1.1 to 200 times the melt energy density for SiO2, and develop an approximate relation that gives the mesh resolution needed for a desired percent error in the blow-off momentum as a function of deposited energy density and melt depth. Using these mesh constraints, the response of our one-dimensional SiO2 surface to the energy sources is simulated, and lower bounds are placed on the melt/vapor blow-off momentum as a function of deposited energy density and source energy type.

Simulation of response functions of fast neutron sensors and development of thin neutron silicon sensor

On radiation detection using silicon sensor, signals are produced from collected charges in a depletion layer; however, for high-energy particles, this depletion layer is extended due to funnelling phenomenon. The lengths of charge collection were experimentally obtained from proton peak energies in measured pulse-heights. The length is extended with increasing proton energy of up to 6 MeV, and then, is constant over 6 MeV. The response functions of fast neutron sensors were simulated for 5- and 15-MeV monoenergetic and (252)Cf neutron sources using the Monte Carlo N-Particle eXtended code. The simulation results agree well with the experimental ones, including the effect of funnelling phenomenon. In addition, a thin silicon sensor was developed for a new real-time personal neutron dosemeter. Photon sensitivity is vanishingly smaller than neutron one by a factor of 5×10(-4).

Measurement of Cross-sections of Yttrium (n,xn) Threshold Reactions by Means of Gamma Spectroscopy

Neutron activation and gamma spectrometry are usable also for the determination of cross-sections of different neutron reactions. We have studied the cross-section of yttrium (n,xn) threshold reactions using quasi mono-energetic neutron source based on the reaction on 7Li target at Nuclear Physics Institute of ASCR in Rez. Yttrium (n,xn) threshold reactions are suitable candidates for fast neutron field measurement by activation detectors. Fast neutron field monitoring is necessary already today at a wide range of accelerator facilities and will gain on importance in future fast reactors of generation IV, accelerator transmutation systems or fusion reactors. The knowledge of the cross-sections is crucial for such purpose. Unfortunately, the cross-section is sufficiently known only for 89Y(n,2n)88Y reaction. For higher orders of reactions there are almost no experimental data. Special attention was paid to the 89Y(n,3n)87Y reaction. The nuclei are produced, both in the ground state with half-life 79.8hours and in the isomeric state with half-life 13.38hours. The isomer decays mainly through the gamma transition to the ground state, the beta decay of the excited state is negligible within our accuracy. The cross-sections of both 87Y productions were analyzed.

Studies of (n,xn) cross-sections in Al, Au, Bi, Cu, Fe, I, In, Mg, Ni, Ta, Y, and Zn by the activation method

The article summarizes series of experiments devoted to studies of (n,xn) cross-sections by the activation method and states numerical values of measured cross-sections. The activation samples are used to measure neutron flux produced during accelerator driven system studies. The threshold (n,xn) reactions are used but experimental cross-section data of such reactions are mostly not available for neutron energies above 30MeV. We focused for this reason on activation materials (Al, Au, Bi, In Ta, and Y) commonly used for such purposes and we also measured other materials (Cu, Fe, I, Mg, Ni, and Zn). The cross-sections were studied using quasi mono-energetic neutron sources based on proton reactions in a 7Li target in the energy range from 17MeV up to 94MeV. We observed a good agreement of the obtained data with the experimental data in the EXFOR database and also with the calculations performed using the code TALYS 1.4. Many of presented measurements represent the first ever measurement for a given reaction at a given neutron energy.

Comparison of unfolding approaches for monoenergetic and continuous fast-neutron energy spectra

Abstract Proton-recoil detectors offer the possibility to unfold fast-neutron energy spectra of various sources. However, quantifying the confidence of the unfolding methodology is a complex task. In this paper, we present a comparative analysis of the maximum-likelihood, expectation-maximization (MLEM) method and one-step-late (OSL) method for neutron energy spectra unfolding. The analysis is performed on Monte Carlo simulated data for several monoenergetic neutron sources and continuous-in-energy 252 Cf, 241 Am–Be and 241 Am–Li neutron sources. The results obtained for the monoenergetic neutron spectra show that both unfolding methods provide results that are in good agreement with the reference data. Very good agreement between the unfolded and the reference data is achieved for 252 Cf, 241 Am–Be, and 241 Am–Li neutron spectra by using the OSL method. In the paper it is demonstrated that the MLEM and OSL methods can be applied to accurately unfold the simulated pulse-height distributions for organic liquid scintillation detectors. Comparative analysis between the two unfolding methods has shown that the OSL method has superior unfolding performance than the MLEM method.

Reference neutron fields of the KIGAM for the neutron energy range between 144 keV and 2.5 MeV

The Korea Institute of Geoscience and Mineral Resources (KIGAM) is constructing a reference neutron field facility as a national project. Neutron fields consist of mono-energetic sources of 144 keV, 250 keV, 565 keV, and 2.5 MeV have a fluence range from 102 neutrons/cm2/sec to 103 neutrons/cm2/sec. The systems for the reference neutron fields, such as a duo-plasmatron ion source, a 4-MHz beam bunching system, a neutron chamber, an irradiation room, a neutron time-of-flight (n-TOF) system, a long-counter, and a sample moving system, were designed and fabricated. The neutron energies of the reference neutron fields and their spreads were observed by using the n-TOF system. The neutron fluence was measured by using a long-counter for energies below 1 MeV and a proton-recoil counter for 2.5 MeV. The long-counter efficiency was calibrated by the Japan Atomic Energy Agency (JAEA) which had a traceability of mono-energetic neutron sources to both Japanese and international standards. The efficiency of the proton-recoil counter was obtained by using a calculation with detailed construction information.

Combining GEANT4 and TIARA for Neutron Soft Error-Rate Prediction of 65 nm Flip-Flops

Experimental neutron characterizations of standard and radiation-hardened-by-design (RHBD) flip-flops (FFs) are compared to combined GEANT4 and TIARA simulations. Good agreement is found between experiment and simulation for both architectures exhibiting soft error-rate (SER) values ranging over more than two decades. In-depth analysis of TIARA simulation results demonstrates that different underlying mechanisms are at the origin of observed soft errors: neutron-silicon (n-Si) elastic scattering for standard 65 nm FF while multiproduct nuclear reactions are for the RHBD structure. Finally, TIARA simulation results for monoenergetic neutron sources and atmospheric-like (i.e., synthetic) neutron spectra are discussed together with their impact on SER.

Magnet Design of 4 MeV Cyclotron for the BNCT

Research concern about neutron radiography has been increasing in Korea since 2001. Neutron radiography can help for the weakness of X-ray image. Neutron radiography can be used in not only non-destructive inspection but also basic science research. For those reasons more compact mono-energetic neutron source is needed. The 4 MeV deuteron cyclotron can be a candidate for the production of mono-energetic neutrons. In this paper optimal 4 MeV deuteron cyclotron magnet design is shown for neutron generation by d(d, n)He reaction. All field simulations have been performed using POSSION and OPERA-3D (TOSCA) for magnet system design. The required beam current is 1 mA, 4 MeV. The first purpose of this cyclotron is to generate fast neutrons with deuterium gas target. The second purpose of this cyclotron is material science and nuclear technology research in biology, physics and chemistry. Also it will be used in industries including neutron radiography.

Response function of a superheated drop neutron monitor with lead shell in the thermal to 400-MeV energy range

Superheated drop detectors are currently used for personal and environmental dosimetry and their characteristics such as response to neutrons and temperature dependency are well known. A new bubble counter based on the superheated drop technology has been developed by Framework Scientific. However, the response of this detector with the lead shell is not clear especially above several tens of MeV. In this study, the response has been measured with quasi-monoenergetic and monoenergetic neutron sources with and without a lead shell. The experimental results were compared with the results of the Monte Carlo calculations using the 'Event Generator Mode' in the PHITS code with the JENDL-HE/2007 data library to clarify the response of this detector with a lead shell in the entire energy range.

A Monte Carlo simulation of monoenergetic neutrons traversing rectangular and spherical polyethylene moderators

A Monte Carlo-based simulation of the transport of a series of monoenergetic neutron sources through first a rectangular block of 0.93 g cm(-3) density polyethylene and secondly through a sphere made of the same substance is presented here. In both instances, the neutron fields are monitored at closely spread intervals through the moderator mass, producing a lot of data in the process. To reduce the amount of data presented, a figure of merit is created by estimating the cross section for each discrete neutron energy and by applying this to the number of neutrons present of each energy giving an arbitrary response figure. This work was undertaken in order to aid the design and development of a novel neutron spectrometer.

Investigation of fast neutron shielding characteristics depending on boron percentages of MgB2, NaBH4 and KBH4

The macroscopic cross-section Σ and average neutron fluence in matter Φ are usable factors to comment neutron shielding property of samples. In this paper, we have used MgB2, NaBH4 and KBH4 samples including different percentages of boron. Neutron macroscopic cross-section measurements of them have been done by using a source of mono-energetic neutrons (E eff = 4.5 MeV 241Am–Be). Average neutron fluence values and double differential fast neutron flux distributions of each samples calculated by using FLUKA Monte Carlo code. Also half value layers (HVLs) of samples are compared to paraffin which is one of the most neutron moderators. As a result, growing boron concentration can raise neutron shielding property of materials.

Neutron dose transmission measurements for several new concrete samples including colemanite

We have prepared four baryte and four concrete samples having respectively 0%, 5%, 10% and 15% colemanite concentrations. Neutron dose transmission measurements have been done by using a source of mono-energetic neutrons ( E eff = 4.5 MeV– 241Am–Be). It has been shown that when colemanite percentages of the samples increase, neutron dose transmission values for the samples decrease. It is thus possible to enhance the neutron shielding property of baryte and ordinary concrete by adding different proportions of colemanite.

Mono-Energetic Neutron Source by 7Li(p,n)7Be Reaction

Mono-Energetic Neutron Source by 7Li(p,n)7Be Reaction

Evaluation of D(d,n) 3 He reaction neutron source models for BNCT irradiation system design

A mathematical method was developed to calculate the yield, energy spectrum and angular distribution of neutrons from D(d,n) 3 He (D-D) reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV. The data of energy spectrum and angular distribution were applied to set up the neutron source model for the beam-shaping-assembly (BSA) design of Boron-Neutron-Capture-Therapy (BNCT) using MCNP-4C code. Three cases of D-D neutron source corresponding to incident deuteron energy of 1000, 400 and 150 keV were investigated. The neutron beam characteristics were compared with the model of a 2.45 MeV mono-energetic and isotropic neutron source using an example BSA designed for BNCT irradiation. The results show significant differences in the neutron beam characteristics, particularly the fast neutron component and fast neutron dose in air, between the non-isotropic neutron source model and the 2.5 MeV mono-energetic and isotropic neutron source model.

Single Event Upsets Induced by 1–10 MeV Neutrons in Static-RAMs Using Mono-Energetic Neutron Sources

The neutron-induced SEU sensitivity in the 1-10 MeV energy range is investigated using monoenergetic neutron beams at 2.5, 4, 6, and 14 MeV. Below the 0.25 mum technology node, bulk technologies exhibit a relatively high sensitivity to neutrons between 4 and 6 MeV which is explained by the contribution of alpha particles coming from (n, alpha) reactions. In the terrestrial environment, the contribution to SER of neutrons in this energy range exceeds 10%.

A Novel Feature of Neutron-Induced Multi-Cell Upsets in 130 and 180 nm SRAMs

Bit-multiplicity of neutron-induced single event upsets (SEU) in CMOS SRAMs formed with 130 and 180 nm technologies was analyzed using mono-energetic, quasi-mono-energetic, and spallation neutron sources in various accelerator facilities. The energy dependence of the ratio of multi-cell upsets (MCUs) to the total number of upsets can be described by a Weibull-type function with the threshold energy of the MCU. The 130 nm SRAMs show a novel feature of multi-cell upsets (MCUs) including frequency distribution of the multiplicity of error bits. In the case of the 130 nm SRAM, the probability function of the MCU can be expressed as a sum of exponential and Lorentzian functions of the multiplicity of error bits. According to previous results of 3-dimensional device simulation, the Lorentzian component can be due to bipolar action.

Development of advanced-type multi-functional electronic personal dosemeter

An advanced-type small, light, multi-functional electronic personal dosemeter has been developed using silicon semiconductor radiation detectors for dose management of workers at nuclear power plants and accelerator facilities. This dosemeter is 62 x 82 x 27 mm(3) in size and approximately 130 g in weight, which is capable of measuring personal gamma ray and neutron dose equivalents, Hp(10), simultaneously. The neutron dose equivalent can be obtained using two types of silicon semiconductors: a slow-neutron sensor (<1 MeV) and a fast-neutron sensor (>1 MeV). The slow neutron sensor is a 10 x 10 mm(2) p-type silicon on which a natural boron layer is deposited around an aluminium electrode. The fast neutron sensor is also a 10 x 10 mm(2) p-type silicon crystal on which an amorphous silicon hydride is deposited. The neutron energy response corresponding to the fluence-to-dose-equivalent conversion coefficient given by ICRP Publication 74 has been evaluated using a monoenergetic neutron source from 250 keV to 15 MeV at the Fast Neutron Laboratory of Tohoku University. As the result, the Hp(10) response to neutrons in the energy range of 250 keV and 4.4 MeV within +/-50% difference has been obtained.