The calibration procedure of silver activation DETECTORs using a mono energetic NEUTRON source with 5MeV beam energy at FRS of JAERI (Japan Atomic Energy Research Institute) is described in this paper. These DETECTORs are fabricated for measuring fusion NEUTRON burst emitted of z-pinch and plasma focus devices in ASRL (the Atomic Science Research Laboratory).

In this study, calibration process was carried out for deigned new CR-39 nuclear track DETECTOR for protons, NEUTRONs and alpha particles separately under the same etching condition. In order to aim this purpose, americium-beryllium standard source (241Am-Be) and Plexiglas phantom for NEUTRON irradiation, brass collimators and americium standard source (241Am) for alpha irradiation and the accelerator and Van de Graff accelerator for proton irradiation were employed. Sodium hydroxide solution 6. 25N at the temperature of 85 ° C was used for CR-39 etching. Considering obtained results, different detection shields were designed to distinguish between fast NEUTRON particles, thermal NEUTRONs, albedo NEUTRONs, protons and alpha in mixed radiation fields. Moreover, both the contribution of each particle and the ability of the designed DETECTOR to discriminate energy of charged particles were found.

The effect of the host medium viscosity of the superheated drop DETECTOR on formation and stability of Freon-12 bubbles produced by NEUTRON irradiation has been investigated. By variation of the acrylamide and methylenbisacrylamide concentrations on monomer solution, different polyacrylamide gels with various viscosities of 1 up to 13 Pa-s were prepared. The gels were utilized to prepare the superheated droplet DETECTOR. The irradiation results of the prepared superheated droplet DETECTOR using an 241Am-Be NEUTRON source and the 2.89 MeV NEUTRONs obtained from the d-d NEUTRON generator showed that the formed bubbles in the fabricated DETECTOR, based on non-crosslinked polyacrylamide gel, with the viscosity of 5-6 Pa-s were stable and could be counted after irradiation with the naked eye. The number of bubbles were found to be proportional to the NEUTRON fluence.

Fast NEUTRON flux (14.8 MeV) of a NEUTRON generator has been measured by activation technique. The measurements performed using Cu and Ni threshold DETECTORs. 62CU and s7Ni were produced through 63CU (n, 2n) 62Cu and 58ni (n, 2n) 57Ni reactions. They decay by emitting 511 keV and 1377 keV gamma rays, respectively. The half life of 62CU is 9.74 min and that of 57Ni is 36 hours. The flux of NEUTRON has been calculated by measuring the activity after the irradiation time. Gamma spectroscopy of the activated foils was performed using a HPGe DETECTOR. By employing this technique the NEUTRON flux of 2.64x107±3% n/s was obtained for 60mA deuteron of 110keV energy, bombarding a solid target of 3H.

Background: Miniature NEUTRON sources with high NEUTRON flux have abundant applications in medicine, industry and researches. The most important general characteristic of miniature NEUTRON sources is their diameter which is 3mm in average. In this research, we have surveyed and designed an Am-Be miniature NEUTRON source fabrication. Materials and Methods: This investigation resulted in creation of an Am-Be NEUTRON source, using beryllium metal powder with 98% carat and 100-200mm mesh and Americium source with activity of about 200mCi. NEUTRON source designing was performed under safety and protective factors. The system was designed in two different forms based on the fluent yield of NEUTRON or cut off NEUTRON yield. Results: The mean NEUTRON flux of miniature NEUTRON source was measured as 1.14 (n/sec.cm2), and it was calculated as 2.56 (n/sec.cm2) by MCNP (4C) code. Due to purity and mesh of beryllium, which were not calculated by MCNP code, the calculated flux via Monte Carlo method was approximately 2 times larger than NEUTRON flux from fabricated miniature NEUTRON sources. Conclusion: In order to fabricate the miniature NEUTRON sources Am-Be with high efficiency, the americium sources with high activity and the target material (Be) in different forms are required.

The amount of cosmic rays varies widely with the altitude, latitude and longitude in each region. In this study, the radiation doses due to the cosmic rays were estimated in two steps: in the first step, the NEUTRON and gamma components of the radiation dose were measured for a roundtrip flight on 3 flight routes (Shiraz-Asaluye, Asaluye-Rasht and Shiraz-Mashhad) using a gamma-tracer photon DETECTOR and a Thyac 190N, NEUTRON DETECTOR. The minimum values of the measured gamma and NEUTRON doses of 0.15 and 0.04 mSv were measured on the Asaluyeh-Shiraz route at the lowest altitude of 19000 ft, while for Rasht-Asaluyeh route at an altitude of 35000ft those values were found to be 2.52 and 1.09mSv, respectively. In the second step, a number of aircrew members were equipped with thermoluminescence dosimeters (TLD cards) for evaluating the gamma dose and polycarbonate dosimeters (SSNTD) for assessing the NEUTRON dose for one year. The measured value of the annual effective dose received by the crew ranged between 0.5 mSv/y and 1.16 mSv/y, with an average of 0.9 mSv/y for the gamma component and between 0.37 mSv/y and 0.77 mSv/y with an average of 0.61 mSv/y for the NEUTRON component. The results of this investigation are comparable with the investigations that have been conducted in other countries. For instance in UK, the reported annual effective dose of aircrew is about 2mSv, and in Canada, it is estimated to be between 1 to 5mSv, depending on the flight situations (such as the latitude and longitude of the cities, the flight altitude, etc).

In this paper, the fast NEUTRON spectra at altitudes of 3 and 5 km were unfolded by the response of Superheated Drop DETECTORs and using an Adaptive Network-based Fuzzy Inference System (ANFIS). ANFIS is a Takagi-Sugeno Fuzzy Inference System implemented in the framework of adaptive networks. This tool works similarly to human thinking in dealing with uncertain and erroneous problems. The response matrix of five Superheated Drop DETECTORs under various external pressures was calculated by an application developed using the Geant4 simulation toolkit and was used to obtain inputs of ANFIS. Also, the NEUTRON spectra of the IAEA technical reports were utilized as the targets. The reference spectra were unfolded with RMSEs of 0. 005 and 0. 011. The relative agreement between the unfolded and reference spectra shows that these DETECTORs and ANFIS can be used as a new technique for unfolding NEUTRON spectra produced by cosmic radiations in the atmosphere.

Investigation of Hydrocarbon reservoir is important, so it is essential to predict and explore them precisely. One of the methods is well logging, which can transfer the probe or tool in the well to measure one or more characteristics. Nuclear well logging includes radioisotope source and at least one DETECTOR. In this work, emission direction of NEUTRONs from the 241Am-Be NEUTRON source toward the calcite formation has been investigated using MCNPX 2. 6 to obtain the best precision in determining the liquid porosity. The results show that emission direction of 20 in degree with the resolution porosity of 3% in counts is recorded in DETECTORs; as well; there is no decrease in depth penetration, providing speed of tools.

Fast NEUTRONs that are produced via compact NEUTRON generators have been used for thermal and fast NEUTRON radiographies. In order to investigate objects with different sizes and produce radiographs of variable qualities, the proposed facility has been considered with a wide range of values for the parameters characterizing the thermal and fast NEUTRON radiographies. The proposed system is designed according to article 4 of the Restriction of Hazardous Substances Directive 2002/95/EC, hence, excluded the use of cadmium and lead, and has been simulated using the MCNP4B code. The Monte Carlo calculations were carried out using three different NEUTRON sources: deuterium-deuterium, deuterium-tritium, and tritium-tritium NEUTRON generators.

In this paper, the MCNPX code is applied for feasibility study of using the Isfahan MNSR as a NEUTRON source to be used for NEUTRON radiography. To produce a good NEUTRON beam, in terms of intensity and quality, aluminum (Al) with a thickness of 0. 7 cm, bismuth (Bi), and lead (Pb) with a thickness of 1 cm are used as a fast NEUTRON filter, and the gamma filter, respectively. The L/D ratio of the designed NEUTRON radiography facility is 90 and the diverging angle is 2. 1 degree. The thermal NEUTRON flux, the ratio of thermal NEUTRON to gamma dose rate, and the thermal NEUTRON content at the beam exit plane are evaluated to be 1. 47E+05 n/cm2. s, 2. 96E+06 n/cm2. mR, and 92. 5%, respectively. It was realized that if such a thermal NEUTRON beam is built in Isfahan MNSR, many practical and scientific applications of the NR can be realized.