Isotopic Neutron Source Research Articles

NeutralNet: Development and testing of a machine learning solution for pulse shape discrimination

Accurate description of radiation fields containing neutrons continues to be a difficult task to complete. This difficulty arises because of the inherent sensitivity of neutron detectors to other types of radiation, and the ability of neutrons to generate secondary particles producing mixed field environments. This research looks at the development and performance of various machine learning architectures when applied to the task of pulse shape discrimination with liquid scintillators. This work was carried out with a neutron sensitive liquid scintillator, EJ-301, with signals digitized at 3.2 GHz with 12 bits of resolution utilizing a CAEN DT-5743 digitizer. Measurements were artificially reduced in sampling depth and frequency to investigate the importance of these parameters for performance of the machine learning algorithms. Two isotopic neutron source, 238Pu9Be and 241Am9Be, and three photon sources: 24Na, 60 Co, and 137Cs were used for generation of the training and validation sets as well as for energy calibration. The greatest performance was achieved with a lightly altered implementation of GoogLeNet and with the full sampling rate and bit depth afforded by the digitizer. A true positive rate of 69.17 % was achieved while correctly rejecting 99.9999 % of photon events. This performance ranges from 1.30 % at events greater than 3 MeVee to 89.96 % between 340–1000 keVee. For neutron events with energy below 200 keVee 21.48 % of the validation neutrons are still identified as neutrons.

Remarkable reduction in thermal contact resistance between target and heat exchanger using TIMs in isotope neutron source accelerator

The high thermal contact resistance (TCR) between the target and the heat exchanger in the Isotope Neutron Source Accelerator will cause the high-energy charged particle beam to stay in the target, resulting in a remarkable increase of temperature which affects the production of stable isotopes. At present, the usage of thermal interface materials (TIMs) can fill the gap at the contact surface and thus reduces the TCR. In this paper, aiming at the actual working condition of neutron source, the steady-state TCR measurement experiment is designed under vacuum and high temperature environment. The TCR at the contact surface was reduced after adding graphite films, thermal greases, liquid metals and preloads. The improvement of TCR of the contact surface between palladium (Pd), copper (Cu) and graphite (Gr) were measured respectively. The results show that the TCR between palladium and graphite is reduced by 95 % by adding liquid metal (Gallium alloy). The TCR between copper and graphite was reduced by 91 % by brazing method. Compared with other TIMs, graphite film has the advantages of high thermal conductivity, high temperature resistance and no fluidity. Adding 0.08 mm thick graphite film between Pd/Gr and Cu/Gr contact surface can reduce the TCR by 50 % and 70 %, respectively. The improvement effect between the two contact surfaces was not obvious when 15.9–318.3 kPa preload was applied.

Comparison of H*(10) estimations, due to neutrons and γ-rays, around FANT with two 241Am/9Be sources

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, which has been developed by the Neutron Measurement Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid (LMN-UPM). In FANT, an isotopic neutron source (241Am/9Be) produces the primary neutrons. The design and facility optimization were carried out by extensive Monte Carlo calculations. In addition, Monte Carlo methods were used to evaluate the facility's performance to produce a constant and uniform thermal neutron field; these results were validated through experimental methods. FANT is designed to have two neutron sources; the objective of this work is to estimate the ambient dose equivalent due to neutrons and gamma-rays by Monte Carlo methods, and to compare these values with measured experimental doses. Thus, the performance of FANT with the two 241Am/9Be sources of LMN-UPM, with regard to the ambient dose equivalent H*(10) produced by both neutrons and photons around the facility, is analyzed in this work. The results are compared with those previously obtained in the framework of the results obtained with the LB6411 device around FANT.

Quantification of silicon concentration in sugarcane ash using the neutron activation technique with an isotopic source

In the realm of agronomic research, the neutron activation technique has gained traction, particularly in regions such as Paraguay, for its capability to deploy an isotopic source for the quantification and identification of agriculturally pertinent minerals. Within the ambit of sugarcane cultivation, both silicon (Si) and silica (SiO2) emerge as paramount, owing to their efficacy in fortifying the sugarcane plant's resistance to phytopathogens and pests, as well as their role in mitigating lodging phenomena. Notably, an uptick in silica accumulation has been correlated with amplified biomass and sucrose production metrics. This underscores the imperative for a meticulous quantification of Si and SiO2 within sugarcane derivatives, catering to both agronomic and industrial directives. Addressing the extant lacuna in regional datasets, this investigation endeavored to elucidate the inherent concentrations of Si and SiO2 within distinct anatomical segments of sugarcane, viz., leaves, entire stems, bark, and bagasse. A total of eight sugarcane cohorts, each encompassing 10 specimens, were subjected to the experimental paradigm. Adhering to a fully randomized design, each cohort was trialed in triplicate. Post desiccation at 105°C, the resultant specimens underwent calcination in a muffle furnace at an approximation of 500°C for a sextet of hours. The experimental variables under scrutiny encompassed mass (in grams), Si concentration (in percentage), and SiO2 content. Subsequent quantifications of Si and SiO2 within the ash specimens were operationalized using a nuclear analytical modality coupled with an isotopic neutron source. Post-acquisition, the data were subjected to a unidirectional analysis of variance, with an ensuing post-hoc DMS comparative analysis. Empirical findings highlighted that ash derived from integral sugarcane stems registered a preeminent mass juxtaposed against other segments. Intriguingly, leaf-derived ash manifested the zenith of Si and SiO2 concentrations, quantified at 24% for Si and 51.5% for SiO2. In summation, the neutron activation technique, undergirded by an isotopic source, has manifested its robustness in pinpointing Si concentrations within sugarcane ash, thereby accentuating the potential of leaf ash as a quintessential fount for silica extraction. Such insights bear salient implications for nations, such as Paraguay, where the evolution of agronomic modalities remains an ongoing endeavor.

Radiation characterizations of two isotopic neutron sources merging in one irradiator for experimental applications in the laboratory

Neutron sources are utilized for different aims, such as studying the material’s internal structure, investigating the materials’ crystal structures, and neutron therapy. Using more than one isotopic neutron source in some aims could be beneficial. Herein, a neutron irradiator consisting of two 241Am-9Be neutron sources has been designed. Radiation characterizations, including the neutron fluence and dose rate, were investigated. The Monte Carlo simulation code Monte Carlo N-particle transport code version 5 (MCNP5) was used to conduct all the required investigations to achieve an optimal design. Two factors were considered; the first was the dose rate due to both neutron and gamma radiations to meet radiation protection requirements; MCNP5 was used to determine the adequate thickness of the shielding material. The second parameter is the neutron fluence. Two irradiation sites have been proposed, one for fast neutron irradiation and the other for thermal neutron irradiation. By utilizing MCNP5, we have determined the radiation characterization at the two sites to be used as the irradiation sites. After the optimal design had been produced, the thermal and the epithermal neutron fluences inside the two irradiation sites were determined experimentally. The experimental results showed a perfect agreement with the simulation ones. The neutron irradiator designed from merged two isotopic neutron sources could be a benefit in the neutron radiation applications in the labs.

Noise experiments in BRAHMMA subcritical system using isotopic Poisson source and accelerator based neutron source

Reactivity measurement using noise methods have been carried out in zero power subcritical system BRAHMMA, installed in BARC, India. During noise experiments, an Am-Be neutron source (1Ci) has been placed at the centre of the subcritical system. However, the proposed external neutron source in ADS, based on particle accelerator, is different from isotopic neutron source and the inherent fluctuations in beam current and accelerating voltage makes the source non-Poisson. In this context, Degweker and Rana had postulated a noise theory based on exponentially correlated Gaussian source characteristics for ADS. In noise experiments based on this postulate, a D-T neutron generator has been used in pulsed mode and source characterisation has been carried out to determine the distribution function, source correlation factor and D+ beam pulse shape. It has been observed that the source is Gaussian in nature, the source correlation factor is very large compared to the prompt neutron decay constant and the pulsed D+ beam is rectangular in shape. In noise experiments, the time stamped data have been analysed using various noise methods. The measured prompt neutron decay constant using Poisson source and correlated Gaussian source are in good agreement with theoretical value and amongst them.

Research on angle sensitivity of the boron-lined multilayer converter neutron detector

In our previous work, a new design for a gaseous neutron detector composed of a boron-lined multilayer converter was proposed to replace the 3He proportional counter, and a prototype detector was constructed to test its performance and characteristics. The neutron detection efficiency varies for different incident angles of neutrons to the converter surface, mainly due to the effective area of the converter and the emitting probability of the alpha and 7Li. Hence, the variation of detection efficiency with the incident angle is analyzed theoretically in this work, and then the GEANT4 simulations and preliminary experimental tests are conducted to verify the calculation results. The highest detection efficiency is obtained at 1.72° (100 mm length, 3 mm gap, and θ = atan (3/100) ≈ 1.72°), the detection efficiency decreases sharply when the angle approaches 0°, and the detection efficiency decreases gradually when the incident angle is larger than 1.72°. The simulation results and experimental results are consistent with the trend of theoretical calculation. Therefore, the proposed detector is suitable for angle-sensitive applications, such as small angle neutron scattering and locating the isotopic neutron sources.

The calculation of self-shielding correction factors for large samples in 241Am–Be isotopic neutron source

In the present study, it is aimed to calculate neutron self-shielding correction factors for cylinder sample geometry using different methods in the literature in the application of k0-NAA standardization method with 241Am–Be isotopic neutron source. These correction factors are important in calculating the accurate concentration values for large samples with a high neutron absorption cross-section in the application of the k0-NAA method. In order to see the validity of the applied theoretical and experimental methods, the obtained concentration values are given comparatively with respect to the known Mn concentrations. The mean concentration difference ranges from −3.4% to −9.7%. The thermal neutron self-shielding correction depends on all elements in the sample that absorb thermal neutrons. In k0-NAA applications with large samples, although time-consuming, it is recommended to use the experimental methods after calculating the necessary parameters with the help of samples with known composition.

Performance of two magnesium borate compounds, with thermoluminescent properties, to γ-rays and neutrons

There is a permanent effort to search for natural and artificial materials with thermoluminescent properties useful for radiation dosimetry. In the aim to understand the radiation interactions with these materials, several features are determined. Here, the performance to photons and neutrons of MgB4O7:Dy and Mg2B2O5:Dy,Na were studied. The mass interaction coefficients, the linear attenuation coefficients, the mean free path and the Zeff for photons were calculated. In addition, the absorbed dose response to gamma rays and neutrons were estimated. The linear attenuation coefficients to gamma-rays for MBO:Dy,Na is larger than the coefficients for MBO:Dy due the density and Dy concentration, however the absorbed dose response for both materials are approximately the same due to the Zeff. The absorbed dose response to monoenergetic neutrons and those from isotopic neutron sources, and with Maxwellian distribution has the same trend. This response is larger for 252Cf/D2O; the absorbed dose response to neutrons is larger for MBO:Dy due to the boron concentration.

SENSITIVITY ANALYSIS OF PWR SPENT FUEL DUE TO MODELLING PARAMETER UNCERTAINTIES USING SURROGATE MODELS

In the BEPU (Best Estimate Plus Uncertainty) framework, uncertainty quantification (UQ) is a requirement to improve confidence and reliability of code predictions. Over the years, a lot of works have been done to quantify uncertainties in code predictions of spent nuclear fuel (SNF) characteristics due to nuclear data uncertainties. The purpose of this study is to quantify the uncertainty in pressurized water reactor (PWR) fuel assembly radiation source terms (isotopic inventory, activity, decay heat, neutron and gamma source) due to uncertainties in modeling parameters. The deterministic code STREAM is used to predict the source terms of a typical PWR fuel assembly following realistic and detailed irradiation history. For the sensitivity analysis (SA) and UQ, surrogate models are developed based on polynomial chaos expansion (PCE) and variance-based global sensitivity indices (i.e., Sobol’ indices) are employed. The global SA identifies the less important uncertain parameters, showing that the number of uncertain input parameters can be reduced. The surrogate model offers a significantly reduced computational burden even with large number of samples required for the SA/UQ of the model response.

Experimental characterization of FANT, a new thermal neutron source

FANT is the acronym of Enhanced Thermal Neutron Source (Fuente Ampliada de Neutrones Térmicos, in Spanish). This is a parallelepiped box of high-density polyethylene moderator and an isotopic neutron source. The moderator has a cylindrical irradiation chamber where a rather uniform thermal neutron flux is obtained. The FANT design was previously optimized and the neutron spectra were estimated by Monte Carlo calculations with the MCNP6.1 code. To check the characteristics of the FANT thermal neutron field, measurements have been performed at the reference point inside the irradiation chamber with a Bonner sphere spectrometer holding a small 6LiI(Eu) thermal neutron detector. To unfold the neutron spectrum BUNKIUT with UTA4 response matrix and NSDann Ver 4.0 codes were used. Some issues have been found and recommendations are made about the use of large BSS inside narrow spaces, and about the capacity of NSDann code to unfold these kind of spectra. However, the results confirm that the moderation process in FANT is very effective and allows obtaining useful thermal neutron fluence rates.

Neutron self-shielding correction in application of k0-NAA standardization method to determine Mn content in manganese ores

In this study, it is aimed to investigate neutron self-shielding effect in application of k0-NAA standardization method to determine Mn contents in manganese ores. Because the neutron self-shielding correction factors are significant factors in application of k0-NAA method for large samples. To estimate these, several Mn-ore samples and Au monitor were irradiated separately in a 37 GBq 241Am–Be neutron source irradiation unit and then they were measured by using a 78.5% relative efficient HPGe detector. In this study, the sigmoid function approaches for calculation of Gth and Ge factors were applied for the right cylinder sample geometry at known elemental compositions in samples, determined by EDXRF technique with a 25 mm2 Silicon Drift Detector (SDD). The present results indicate that k0-NAA method is applicable successfully to determine Mn contents in Mn-ore samples providing that Gth and Ge factors are estimated accurately even if a low-flux isotopic neutron source is used.

Capacity Building in Nuclear Science and Technologyin the Philippines through the Use and Operation of SmallNeutron Sources for Education, Training, and Research

For the past three decades, the Philippines’ expertise in nuclear science and technology (S&T) has diminished resulting from the shutdown of the Philippine Research Reactor-1 (PRR-1) in the 1980s. In addition, the mothballed Bataan Nuclear Power Plant (BNPP) and the low confidence in nuclear technologies led to the non-prioritization of nuclear science in the country. Consequently, new nuclear facilities were never re-introduced, and the transfer of knowledge declined and nuclear applications were limited to radiation and isotopic studies. If the current initiatives of the Philippines in the use of nuclear energy will push through, the country will need competent human resources who will be responsible for building the nuclear facilities and their safe operation. To augment the declined nuclear expertise in the country, the DOST-PNRI implemented capacity building activities to reestablish and sustain knowledge and expertise in nuclear S&T in the country. Activities that were implemented include the development of training materials for undergraduate students and the development of research facility using isotopic neutron sources. This paper discusses the capacity building strategies implemented and their significant outputs, as well as plans for sustainability and continual development.

CHARACTERIZATION OF 241AM-BE NEUTRON SOURCE USING THRESHOLD FOIL ACTIVATION TECHNIQUE

The aim of the study is to characterize 241Am-Be isotopic neutron source with 5 Ci activity and nominal neutron yield 1.1×107 n/s. The emitted neutrons were measured employing threshold multi-foil activation technique using the following reactions: 115In(n,n'), 58Ni(n,p), 64Zn(n,p), 27Al(n,p), 56Fe(n,p), 59Co(n,α), 24Mg(n,p), and27Al(n,α). The threshold reactions were selected based on their half-lives, adequate threshold energy and cross-section. The reaction rate of each sample foil is measured by detecting the gamma ray spectra produced as a result of reaction products decay. A hyper pure germanium detector (HPGe) is used for gamma-ray spectrum measurements. The obtained spectra were analyzed using Genie2000 software and the associated system absolute efficiency was calculated using Efftran software. Obtained reaction rates were compared with available published data.

Influence of neutron irradiation and phosphorus on the performance of groundnut [<i>Arachis hypogaea</i> (L.)] genotypes grown on a P-deficient alfisols at Samaru, Nigeria

There is a dearth of information on the effective irradiation dose for inducing variability in groundnut genotypes. A screen house experiment was conducted at the Department of Soil Science, Ahmadu Bello University, Zaria (ABU). It was to investigate the effect of induced genetic variability (by neutron irradiation) and four phosphorus rates (0, 20, 40 and 60 kg P2O5 ha-1) on the performance of some groundnut [(Arachis hypogaea (L.)] genotypes grown on a P-deficient Alfisols at Samaru, Zaria. A fraction of each of the two genotypes (SAMNUT 24 and SAMNUT 26) was irradiated, in a 5-Curie Americium-Beryllium (Am-Be) isotope neutron source, having a thermal neutron flux of 2.7387 x 104 n cm-2 sec-1 for six days, at the Centre for Energy Research and Training, ABU. The irradiated (G1A and G2A) and nonradiated (G1B and G2B) seeds; and the P rates were laid out in a randomised complete block design (RCBD), in a screen house, and replicated thrice. Data collected during the study included; plant height, numbers of branches, flowers, nodules, leaves and chlorophyll content (CC). Both G1A and G2A significantly (P<0.0001) produced the highest CC, highest numbers of branches, leaves, flowers, nodules and effective nodules; and the tallest plants. Treatment with 60 kg P2O5 ha-1 significantly (P<0.0001) increased the number of leaves, flowers, branches and plant height. The irradiated genotypes were generally observed to significantly perform better than their non-radiated counterparts under the application of both 60 kg P2O5 ha-1 and 20 kg P2O5 ha-1. Thus, application of 20 kg P2O5 ha-1 is recommended for optimum plant height, flowers and CC in groundnut production; while 60 kg P2O5 ha-1 is recommended for optimum numbers of branches and leaves, especially when the irradiated SAMNUT 24 and SAMNUT 26 are to be used for production of both seeds and biomass.Keywords: Alfisols, Chlorophyll content, Neutron irradiation, SAMNUT

Role of isomeric state formation on the measurement of thermal neutron cross section and resonance integral

The isomeric thermal neutron cross section and isomeric resonance integral of 109Ag(n, γ)110mAg, 133Cs(n, γ)134mCs and 134gCs, and 136Ba(n, γ)137mBa reactions were investigated together with 115In(n, γ)116mIn monitor reaction. These residual nuclei have broad half-life time scale suitable for our investigation. Moderated neutrons from steady AmBe source were used for activation of samples with natural abundances. Field was monitored and mapped using gold and indium activation. The formulae used for neutron activation analysis were derived with emphasizing the different interpenetrations of cadmium transmission factor and self-shielding factors. The k0-factor for γ-rays in the residual nuclei were measured. The isomeric thermal neutron cross section and resonance integral for 115In(n, γ)116mIn were evaluated to be 162.6 b and 2585 b. These data were used to measure the k0-factors; and compared to reported values to confirm the procedure. The thermal neutron cross section and resonance integral for 133Cs(n, γ)134mCs were found to be 2.64 ± 0.11 b and 42 ± 1.7 b, respectively; while those of the 109Ag(n, γ)110mAg reaction were 4.09 ± 0.35 b and 68 ± 6 b. Thermal neutron cross section for 136Ba(n, γ)137mBa was identified as 0.032 ± 0.003 b, while the resonance integral could not be evaluated as a result of interfering reactions. Model calculations were done using EMPIRE code to simulate isomeric ratio and compared with the experimental results. The feeding of isomeric state from neutron reaction is more sensitive to the variation of neutron flux distribution than the nucleus formation. Steady neutron field could be retained with isotopic neutron source with moderation setup and geometry of suitable homogeneity and isotropy.

Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field

Abstract The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation field. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron·cm−2·s−1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron field (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron flux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fitted into the Cd filter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect.

Neutron Activation Analysis of Some Building Material

Neutron activation analysis (NAA), based on the comparator method, has the potential to fulfill the requirements of a primary ratio method as defined in 1998 by the Comité Consultatif pour la Quantité de Matière — Métrologie en Chimie (CCQM, Consultative Committee on Amount of Substance — Metrology in Chemistry Studies of Neutron Activation Analysis (NAA) have been carried out on different Egyptian building material samples. The technique of neutron activation analysis is based on the measurement of radiation released by the decay of radioactive nuclei formed by neutron irradiation of the material. The most suitable source of neutrons for such an application is usually a research reactor. The samples that can be analyzed with this method stem from a number of different fields, including medicine, nutrition, biology, chemistry, forensics, the environment and mining. Neutron activation analysis can be performed in a variety of ways. This depends on the element and the corresponding radiation levels to be measured, as well as on the nature and the extent of interference from other elements present in the sample. Most of the methods used are non-destructive, based on the detection of gamma radiation emitted by the irradiated material after or during the irradiation. Next to education and training, neutron activation analysis is the most widely used application of research reactors. Almost any reactor operating at 10-30 kilowatt of thermal power is capable of providing a sufficient neutron flux to irradiate samples for selective applications of this analysis technique. Another method of NAA by using two Am-Be isotopic neutron sources of activity 5 Ci were used in this investigation. The accomplished gamma rays were measured using 70 % HPGe spectrometer. This work demand to estimate the elements contained in cement products and its quality control. X-ray Fluorescence (XRF) measurements were done for confirming our results, and for determining the average neutron flux of 3.7× 103 n/cm2sec. The Natural radioactivities of these samples were measured before the analysis to know the background level of 40K, 238U and232Th nuclei. The results investigated that NAA agree with the results of XRF and the world range of the cement concentration of the essential elements Ca, Al, Na, Fe, Mn, V, Sr and Si.

Characterisation of an isotopic neutron source: A comparison of conventional neutron detectors and micro-silica glass bead thermoluminescent detectors

As a result of their thermoluminescent response, low cost commercial glass beads have been demonstrated to offer potential use as radiation dosimeters, providing capability in sensing different types of ionising radiation. With a linear response over a large range of dose and spatial resolution that allows measurements down to the order of 1mm, their performance renders them of interest in situations in which sensitivity, dynamic range, and fine spatial resolution are called for. In the present work, the suitability of glass beads for characterisation of an Americium-Beryllium (241AmBe) neutron source has been assessed. Direct comparison has been made using conventional 3He and boron tri-fluoride neutron detectors as well as Monte Carlo simulation. Good agreement is obtained between the glass beads and gas detectors in terms of general reduction of count rate with distance. Furthermore, the glass beads demonstrate exceptional spatial resolution, leading to the observation of fine detail in the plot of dose versus distance from source. Fine resolution peaks arising in the measured plots, also present in simulations, are interesting features which based on our best knowledge have previously not been reported. The features are reproduced in both experiment and simulation but we do not have a firm reason for their origin. Of greater clarity is that the glass beads have considerable potential for use in high spatial resolution neutron field characterisation, subject to the availability of a suitable automated TLD reader.

Radiation detector based on 4H-SiC used for thermal neutron detection

In this work we have focused on detection of thermal neutrons generated by 239Pu–Be isotopic neutron source. A high quality liquid phase epitaxial layer of 4H-SiC was used as a detection region. The thickness of the layer was 70 μ m and the diameter of circular Au/Ni Schottky contact was 4.5 mm. Around the Schottky contact two guard rings were created. The detector structure was first examined as a detector of protons and alpha particles for energy calibration. Monoenergetic protons of energies from 300 keV up to 1.9 MeV were used for detector energy calibration and a good linearity was observed. The energy resolution of 35 keV was obtained for 1.9 MeV protons. The 6LiF conversion layer was applied on the detector Schottky contact. In the experiment we used different thicknesses of conversion layers from 5 μ m up to 35 μ m. Measured detected spectra show two parts corresponding to alpha particles detection in lower energy channels and 3H in higher energy channels. We have also performed simulations of thermal neutron detection using MCNPX (Monte Carlo N-particle eXtended) code. The detection efficiency and the detector response to thermal neutrons was calculated with respect to the 6LiF layer thickness. The detection efficiency calculation is found to be in good agreement with the experiment.