Background: The trabecular bone changes in the tibia of C3H/HeN mice were measured 12 weeks after whole body irradiation with various doses of Fast Neutrons (0-2.4 Gy) or 137Cs-generated gamma-rays (0-6 Gy).Materials and Methods: Serum calcium, phosphorus, estradiol concentration and alkaline phosphatase activity were measured. Tibiae were analyzed using microcomputed tomography. Biomechanical property and osteoclast surface level were measured.Results: There was a significant relationship between the loss of bone architecture and the radiation dose, and the best-fittng dose-response curves were linear-quadratic. Mean relative biological effectiveness (RBE) values (Ref. gamma) of 2.05 and 2.33 were estimated for Fast neutron irradiation in trabecular bone volume fraction and bone mineral density, respectively. There was a substantial reduction in osteoclast surface level in tartrate-resistant acid phosphatase-stained histological sections of tibial metaphyses in irradiated mice with high dose of Neutrons.Conclusion: There was a significant relationship between the loss of bone architecture and the radiation dose. The difference of osteoclastic bone resorption may represent a contributor to the low RBE in high dose of irradiation level relative to that of low dose level.

In this paper two geometries for pulsed neutron counter structure have been introduced and to increase the activation counter efficiency, plastic scintillation along with silver foils was used. Cubic and cylindrical geometries for activation counter cell were modeled using MCNP4C code. In respect of absorpstion reaction rate in silver, the number of silver foils and the length of the counter were optimized. The optimum length of 14 centimeters had been proposed for counter cell and because of the economic aspects; the optimum number of silver foils for cubic and cylindrical geometries are 20 and 10, respectively. The optimum data were used to construct a cubic counter and the neutron yield of SBUPF1 plasma focus device was measured by this counter. Experimental results show that about 3.71x107 Neutrons are produced per pulse.

Introduction Electron linear accelerator (LINAC) can be used for neutron production in Boron Neutron Capture Therapy (BNCT). BNCT is an external radiotherapeutic method for the treatment of some cancers. In this study, Varian 2300 C/D LINAC was simulated as an electron accelerator-based photoneutron source to provide a suitable neutron flux for BNCT. Materials and Methods Photoneutron sources were simulated, using MCNPX Monte Carlo code. In this study, a 20 MeV LINAC was utilized for electron-photon reactions. After the evaluation of cross-sections and threshold energies, lead (Pb), uranium (U) and beryllium deuteride (BeD2) were selected as photoneutron sources. Results According to the simulation results, optimized photoneutron sources with a compact volume and photoneutron yields of 107, 108 and 109 (n. cm-2. s-1) were obtained for Pb, U and BeD2 composites. Also, photoNeutrons increased by using enriched U (10-60%) as an electron accelerator-based photoneutron source. Conclusion Optimized photoneutron sources were obtained with compact sizes of 107, 108 and 109 (n. cm-2. s-1), respectively. These fluxs can be applied for BNCT by decelerating Fast Neutrons and using a suitable beam-shaping assembly, surrounding electron-photon and photoneutron sources.

This study investigates the direct effect of Fast Neutrons with the energy ranging from 2MeV to 14MeV, and calculates the single-strand break and double-strand break on the Deoxyribon Nucleic Acid (DNA) atomic structure, using Monte Carlo method. To this end, Geant4 toolkit and its low energy extension, known as Geant4-DNA, were used. The DNA atomic structure extracted from the Protein data bank and water was selected as a substance for the biological matter. The step length in low energy extension works is in the range of nanometer and less. On the other hand, the average free paths of Neutrons in the energy rang from 2MeV to 14MeV was obtained in the unit of centimeters. Under these circumstances, running the program using a computing system will also be lengthy. As a result, the spectrum of secondary particles from neutron interactions with the atoms of water molecules was targeted. The Evaluated Nuclear Data File (ENDF) and the theoretical calculation were used to extract secondary particle spectra. This method reduces the execution time to more than about one-tenth. Then, the relative biological effectiveness (RBE) of the Neutrons were also simulated using 60Co γ-rays as the reference quality. The model succeeded in reproducing the general behavior of RBE as a function of neutron energy, which agrees well with the data reported in the literature.

Neutron filter crystals are widely used in the nuclear industry, particularly in research reactors. One of the most important crystals used is sapphire crystal, which is grown in several countries to ensure good quality. This research aims to compare the neutron performance of sapphire crystals made in Iran with those of externally supplied crystals. An Am-Be source was used in this study to assess the ability of these sapphire crystals to remove Fast Neutrons. Additionally, a monochromatic neutron beam was used to study the impact of these crystals on reducing the intensity of the neutron flux. To verify the accuracy of computational codes in simulating the neutron behavior of these filter crystals, the McStas code was utilized. The results of this simulation were then compared with available experimental data. Comparing the experimental results with those published worldwide revealed that the domestically grown crystals performed comparably to the foreign samples. The experimental findings showed that a 2.5 cm sapphire crystal filtered the monochromatic neutron beam with an energy of 0.06 eV by only about 5.5%, and Fast Neutrons with an energy of more than 200 keV by about 33%. Increasing the crystal thickness to approximately 6 cm increased the filtration of Fast Neutrons to about 60% and monochromatic Neutrons to about 16.5%. Overall, the results of this study indicated that the discrepancies between the simulation and experimental values may be attributed to the lack of detailed information necessary for crystal modeling in the computational code.

Introduction: Low-density bulk metallic glass (BMG) with good structural characteristics has the potential of being used for structural radiation shielding purposes. This study was conducted on two new low-density titanium (Ti)-based BMGs (i. e., Ti32. 8Zr30. 2Ni5. 3Cu9Be22. 7 and Ti31. 9Zr33. 4Fe4Cu8. 7Be22) to investigate their photon and Fast neutron shielding capacities. Material and Methods: The mass attenuation coefficients, half-value layers, effective atomic numbers, and exposure buildup factors of the two BMGs were calculated at the photon energy values of 15 keV and 15 MeV. Computation of mass attenuation coefficients and effective atomic numbers was accomplished using the XCOM and auto-Zeff software, respectively. In addition, the geometric progression procedure-based computer code EXABCal was used for calculating the exposure buildup factors of BMG. The Fast neutron removal cross-sections were also calculated for the two BMGs. The calculated photon and Fast neutron shielding parameters for BMGs were compared with those of lead (Pb), heavy concrete, and some recently developed glass shielding materials and then analyzed according to their elemental compositions. Results: The results showed that though Pb had a better photon shielding capacity, Ti-BMG attenuated photons better than heavy concrete. Furthermore, BMG had a higher neutron removal cross-section, compared to heavy concrete and some recently developed glass shielding materials. The neutron removal cross-sections of Ti32. 8Zr30. 2Ni5. 3Cu9Be22. 7 and Ti31. 9Zr33. 4Fe4Cu8. 7Be22 were obtained as 0. 1663 and 0. 1645 cm-1, respectively. Conclusion: his study revealed that Ti-based BMG with high strength and low density have potential applications in high-radiation environments, particularly in nuclear engineering for source and structural shielding

Introduction: The utilization of high-energy photons in the medical linear accelerator can lead to photoneutron production. This study aimed to evaluate the effect of the physical components of the head, including flattening filter (FF) and multileaf collimator (MLC), as well as the dependence of therapeutic field size on the photoneutron spectrum, dose, and flux. Material and Methods: The present study reported the simulation of the fundamental linac head components of the Varian Clinac 2100 performing in X-ray mode with 18 MV energy by the FLUKA code. The percentage depth dose and lateral dose profile were measured using a PTW thimble chamber to ensure the simulation reliability. Results: Photoneutron spectrum analysis indicated that Neutrons with highest relative biological effectiveness were delivered to the phantom surface, and opening the field from 0×0 to 40×40𝑐 𝑚 2 shifted the spectrum by 24. 545% to the higher energies. The target and the vicinity parts played the most prominent roles in neutron contamination. The relationship between the field size and the photoneutron dose was non-linear, and it reached a peak of 20×20 cm2. Although using small fields formed by the MLC contribute to a lower dose compared to those shaped by the jaws, MLC-equipped machines result in 21. 98% higher dose. Moreover, the flattening filter removal unexpectedly increased the isocenter photoneutron dose by 11. 63%. This undesirable dose can be up to 2. 54 mSv/Gy for the reference field at the isocenter while the out-of-field dose is about 0. 5 mSv/Gy for most of the field dimensions. Conclusion: As a result, it is critical to consider this unwanted absorbed dose, which is seriously influenced by the implemented therapeutic conditions.

Interaction of Fast Neutrons with oxygen atoms of cooling water in reactor core leads to the production of 16N isotope which emits 6.13 MeV gamma ray. A system is developed to measure the16N gamma ray activity using a ф5.08cm x5.08cm NaI(TI) detector. A suitable electronic system is used to count 16N gamma rays in different reactor powers in order to calibrate the 16N system in terms of reactor power. Some non-linearity in the system behavior have been observed and discussions have been presented for this effect.

In this paper, in order to investigate the surface fusion phenomenon in an industrial neutron generator, a solid target with cooling capability was designed and constructed. The first step to achieving this goal is to thoroughly investigate the material and thickness of the layers and substrates suitable for use as solid targets for industrial neutron generators, using SRIM-code simulations. Then, using the simulation results, samples of the solid target were constructed by the sputtering coating method. In addition, due to the importance of the target temperature and its effect on surface fusion, the cooling system using COMSOL multiple physics simulation software, was designed and built. In addition, to insulate the high voltage applied to the target which is in contact with the cooling system, various electrical insulators were studied and suitable insulation was selected, designed, and manufactured. Then, to test the solid targets and their side parts, a suitable vacuum system was designed and constructed. Finally, after designing and constructing all the parts, the system was assembled and set up for final testing. In deuterium filling gas tests, the neutron flux was measured using the LB6411, 3He detector. At around 25 kV voltage and 20 mA current, we were able to detect Neutrons with the rate of 6 × 105 n/s, which was a sign of success. This amount of neutron production indicates duplication of the neutron rate produced by the surface fusion phenomenon.