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.

The aim of this paper is to determine the Fast Neutrons flux incident upon the HPGe detector during the BN thick target gamma-ray yield measurements. These measurements were conducted using the deuteron beam of 0.6, 0.8, 1 and 1.2 MeV. In this work, we measured the Fast Neutrons flux using the 596 keV gamma peak originating from the 74Ge(n,nγ)74Ge and 73Ge(n,γ)74Ge reactions. The contributions of the capture and inelastic scattering reaction over the 596 keV gamma peak yield were measured. The advantage of the method is that no additional equipment is required and the flux of Fast Neutrons is measured online using the collected gamma spectrum.

A thermal Neutron radiography unit using the Neutrons which emits a 10 MeV electron linac compact has been designed and simulated via MCNPX Monte Carlo code. The facility was carried out for an extensive range of values for the collimator ratio L/D, the main parameter which describes the quality of the produced radiographic images. The results show that the presented facility provides high thermal Neutron flux; while with the use of single sapphire Filter fulfills all the suggested values which characterize a high quality thermal Neutron radiography system. A comparison with other similar facilities indicates that the use of a photoNeutron source using a 10 MeV electrons beam is a useful substitutional for radiographic purposes.

Introduction: This study aimed to measure the Neutron contamination of flattening Filter (FF) and flattening Filter-free (FFF) 10-MV photon beams delivered by the Elekta InfinityTM accelerator. Material and Methods: The photoNeutron spectrum produced by the Linac head was evaluated using a Monte Carlo (MC) simulation. The geometry and composition of the head Linac material were modelled based on information obtained from the manufacturer. In this simulation, MC N-Particle Transport Code software (MCNP6) was utilized to model the Linac head and simulate the particle transport. Evaluation of Neutron contamination was carried out for the Linac with FF and without it (i. e., FFF). In this regard, the FFF beam was built by removing the FF from the Linac components. The scoring plane, as the Neutron spectra calculation area for FF and FFF beams, was placed 99 cm from the target. Results: The Neutron type produced by the head Linac Elekta InfinityTM 10-MV photon mode was mostly thermal and Fast. Although there were differences in the Neutron intensity of FF and FFF beams, the type of Neutrons produced by these two modes had the same energy. Based on the photoNeutron reaction energy threshold, it can be concluded that the Neutrons produced from the head Linac were the result of photoNeutron interactions of high-energy photons with molybdenum-96 and tungsten-184 isotopes. Conclusion: The photoNeutron quantity did not change for FF and FFF beams; however, a larger quantity of Neutrons was produced in the FF beam.

Fusion energy due to inertial confinement has progressed in the last few decades. In order to increase energy efficiency in this method various designs have been presented. The standard scheme for direct ignition and Fast ignition fuel targets are considered. Neutrons, electrons and photons transport in targets containing different combinations of Li and be are calculated in both direct and Fast ignition schemes. To compress spherical multilayer targets having fuel in the central part, they are irradiated by laser or heavy ion beams. Neutrons energy deposition in the target is considered using Monte Carlo method code MCNP. A significant amount of Neutrons energy is deposited in the target which resulted in growing fusion reactions rates. It is found that Beryllium compared to Lithium is more important. In an introductory consideration of relativistic electron beam transport into central part of a Fast ignition target, we have calculated electron energy deposition in highly dense D-T fuel and Beryllium layer of the target. It has been concluded that a Fast ignition scheme is preferred to direct ignition because of the absence of hydrodynamic instability.

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

Simulation work provides valuable information on the behavior of different research reactor Neutron analysis facilities. The present study considered Neutron and secondary-gamma dose rate variations by applying a sapphire crystal inside the D channel in Tehran Research Reactor (TRR). The MCNPX computational code was used to model the channel and its designed shield. Neutron and gamma dose rates distributions were calculated with a sapphire crystal modeling to investigate the Neutron diffraction facility hall dose rates. The data from the dose rate simulations were compared with the experimental data available at a power of 4.2 MW from the research reactor. The comparison showed that there is very good conformity between two data series. The simulated Neutron dose rate in front of the main shield overestimated the measurement data by 57% in closed-shutter situation and underestimated the measured data by 32% in open-shutter measurement situation. The investigation has shown that adjusting the crystal size to the channel size is considerably effective, especially at high leakage positions.

Fast Neutron irradiation is one of the most strategic radiation applications of research reactors. Usually, it is performed around the reactor core containing lower Neutron flux. In this paper, a hybrid object has been introduced and analyzed to enhance irradiating applications of the Fast Neutrons in the core of a Material Testing Reactor (MTR). The tool includes an old-type low-consumed HEU control fuel element, a dry channel, and a Cd Filter. It is supposed to be installed at the internal Neutron trap (D4 positions) of TRR core configuration. Calculating results are very promising for using the proposed tool to increase Neutron fluxes, reduce thermal and epi-thermal Neutron fluxes, and shift the Neutron spectrum toward the Fast Neutron region (hardening effect) at the chosen irradiating location. Primary safety parameters are also checked and passed successfully. Furthermore, there are also some other presented safety items which must be checked carefully and conservatively in order to refabricate and install such a irradiating tool in an in-core location of a MTR.

In this paper, we investigate the detection of masked weak moving targets in the adjacent Fast strong target by using FAPC algorithm. The matched Filter of conventional pulse compression radars induces range sidelobes in surrounding a target with high SNR that could mask the smaller targets. The mismatch created in received signal by Doppler phase shift, degrades APC Filter performance in side lobe suppression. In this paper, the FFL-FAPC algorithm is proposed to reduce the range side lobes using the RMMSE estimator in its post-processing method. In various scenarios, we will investigate the detection of masked targets in comparsion with previous methods. Simulation results show that the FFL-FAPC algorithm reduce significantly of computational cost, in addition to provides improved Doppler robustness.