Background: The Poly-Allyl Diglycol Carbonate (PADC) detector is of particular interest for development of a fast NEUTRON dosimeter. Fast NEUTRONs interact with the constituents of the CR-39 detector and produce H, C and O recoils, as well as (n, a) reaction. These NEUTRON- induced charged particles contribute towards the response of CR-39 detectors.Material and Methods: Electrochemical etching was used to enlarge track diameter which was made by low energy recoil protons. Before electrochemical etching, a chemical etching was performed for 1 hour. The responses were also calculated by Monte Carlo simulations, using MCNPX code in different energy bins considering H, C and O recoils. The total registered efficiency and partial contributions of the efficiency, due to interactions with each constituent of CR-39, were calculated.Results: The optimized condition of etchant was obtained to be 6N KOH 15kV.cm-1, and 6 hours etching time. The obtained results show that track efficiency of CR-39 was a function of incident NEUTRON energy. The tracks caused by O and C recoil nuclei were negligible for NEUTRON energies lower than 1 MeV. At NEUTRON energies lower than 1 MeV, only recoil protons would have sufficient energy to leave visible tracks. But, O and C recoils had important contributions in overall response of PADC at NEUTRON energies of few MeV.Conclusion: The efficiency of a CR-39 based dosemeter could be calculated by MCNPX code and the results were in a good agreement with experimental results in energy range of 241Am– Be bare source and 241Am- Be was softened with a spherical polyethylene moderator of radius of 20 cm.

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.

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.

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.

During the cold start-up, the reactor is in sub-critical state. Therefore, the external NEUTRON source cannot be neglected. In this research paper, the analytical solution of NEUTRON point kinetics equations with a group of delayed NEUTRONs in the presence of the pulsed NEUTRON source in a pressurized-water reactor with 235U as a fuel is presented. The analytical solution is based on the expansion of the NEUTRON density in powers of the prompt NEUTRONs generation time. The point kinetics equations with this method are solvable for step and ramp reactivity and lead to better results compared with other analytical works, but are not solvable for sinusoidal reactivity. So, the NEUTRON density response to sinusoidal reactivity is analyzed by using the fixed point and Lyapunov exponents method.

Background: Performing successful BNCT experiments needs a suitable NEUTRON source. Important factors of the NEUTRON beam are flux and energy that are very important in the selection of NEUTRON source. In most centers that use this method for treatment, reactor is a NEUTRON source, which according to characteristics of the reactor appropriated NEUTRONs are very high. High cost of constructing a BNCT center with using of reactor caused seeking other sources such as accelerator indirectly and radioisotope source directly that each has their own advantage and disadvantages. In this paper we created NEUTRON beam by analysis Am-Be NEUTRON source, using NEUTRON filter technique and suitable moderators. The advantages of Am-Be NEUTRON source are being inexpensive, easy portability, small size and well-designed shields. Therefore, by analyzing radioisotope NEUTRON sourcesand Am-Be NEUTRON source specially, we can prepare possible analysis radioisotope NEUTRON source at boron NEUTRON capture therapy. We hope to achieve suitable results by more studies.Methods: NEUTRON beam in 1keV energy created with using Am-Be NEUTRON source and designed suitable NEUTRON filter with using NEUTRON absorbent materials that it will be used in testing BNCT. By studying and Identifying various materials such as oxides Alumina, graphite and beryllium as a moderator and materials such as boron, cadmium and titanium as absorbent materials to a cylindrical crust in filter has been used. NEUTRON Filter has been designed in the investigation of two parts. The first is consisting of a moderator with high scattering and very low percent and it is caused the fast NEUTRON servant brought back his spectrum Am-Be source in this without mono-energetic to the low energy transferred spectrum. Part II filter is consisting of the elements of boron, cadmium and titanium that are absorbent NEUTRON with various energy, therefore they can exchange these NEUTRONs in certain energy to mono-energetic. More analysis, study and designing suitable NEUTRON conductors for increase NEUTRON flux is recommended.Results: NEUTRON filter passes NEUTRON with energy 1keV that can be used in the BNCT experiments. According to data obtained from the implementation MCNP4C code, a peak is obtained in energy 1keV that indicate area under the flux 2.22E-05 n/cm2.s with error 0.0065 for a NEUTRON. Flux obtained can be multiplied at the Am-Be source of power that is equal 108n/cm2.s until the total flux to be achieved. The total flux is obtained 2.22E+03n/cm2.s at 1 cm2. We must multiply total intensity at total area to achieve total NEUTRON flux, Since the flux required for the BNCT experiments is 5*108n/cm2.s with using different ways and designing suitable reflectors and conductors, this NEUTRON flux will be provided.Conclusion: This paper analyzed possible use of radioisotope NEUTRON source by simulation Am-Be NEUTRON source. We can solve many problems that exist for reactor source paying attention to characteristics of radioisotope sources such as being inexpensive, easy portability and small size also more studies are present in this base. Of course, with completing this simulation, we can be hopeful for practicality and remedy of patients in Iran.

Background: NEUTRON contamination in our environment can cause harmful bio-logical effects on human body. Therefore, many efforts have been made to construct a NEUTRON dosimeter to estimate the received dose. Objective: To design a simple NEUTRON dosimeter. Methods: The primary dosimeter had 3He as a spherical thermal NEUTRON detector encircled by paraffin 10 cm in radius. Then, the paraffin sphere was replaced with ICRU that contains soft tissue and dose equivalent determined as a desired output. Finally, an appropriate relation between counts and dose equivalent was found. Results: Results on the energies below 1 MeV demonstrated the similarity of changing process of these two quantities, so they could relate to each other with an adequate factor. To find the best fit, different factors considered and the smallest c2 (goodness of fit) was 1.17×105 At the next step, two covers of cadmium and gado-linium, separately, put around the detector to improve c2, which was 2.51×104 for cadmium cover and 6.33×103 for gadolinium cover. As we see, gadolinium cover fits the curves of counts and equivalent dose in a better way. Conclusion: Applying this simple dosimeter lead us to estimate whole body dose equivalent.