In this paper, the MCNPX code is applied for feasibility study of using the Isfahan MNSR as a neutron source for neutron radiography. To produce a good neutron beam in terms of intensity and quality, the aluminum (Al) with thickness of 0.7 cm, and bismuth (Bi) and lead (Pb) with thickness of 1 cm are used as the 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 2.1degree. 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 1.47E+05 n/cm2.s,  2.96E+06 n/cm2.mR, and 92.5%, respectively. If such thermal neutron beam is built in Isfahan MNSR, many practical and scientific applications of the NR would be realized.

In this study, the neutron and gamma doses in the dry channel and in the internal irradiation site of the Miniature Neutron Source research reactor (MNSR) has been calculated and measured. The MNSR reactor is a light water reactor with a maximum power of 30 kW and equipped with various irradiation facilities, including five irradiated sites, five irradiation sites and a dry channel. The internal irradiation sites have the closest gap to the core of the reactor, with the highest flux and doses available in these locations. Dose calculations have been performed using simulation of the reactor by MCNP computational code and dose measurement using TLD600 and TLD700 thermo-luminescence dosimeters. The experiments have been carried out at both the shutdown and operational status of reactor. In order to validate the computational code, the neutron flux in the internal irradiation site and at the end of the dry channel has been measured by foil activation method and validated by the calculation results. The results of the calculation and measurement of the neutron and gamma doses were in good agreement. The determination of neutron and gamma doses at these sites makes possible such experiments and researches that need to receive a precise amount of neutron and gamma doses.

The use of metal matrix composites can provide a combination of desirable properties of metals as well as the special physical properties of neutron absorber reinforcing particles such as boron carbide, which alone may be brittle. Therefore, in the present study on neutron attenuation power of composite shielding, several Al-B4C composite samples with weight fractions of 5, 10 and 20% B4C have been used. In order to investigate the neutron absorption properties of the studied samples, the MCNP Monte Carlo code and the neutron source of the dry channel of the MNSR reactor with a flux of 2.13E+5 n.cm-2.s-1 have been used  ,which provided in nominal reactor power of 30 kW. The results show that the neutron flux in the presence of 5, 10 and 20% boron carbide samples is predicted to be 1.32E+05 n.cm-2.s-1,1.12E+05 n.cm-2.s-1 and 1.07E+05 n.cm-2.s-1, respectively. With this increase in the percentage of reinforcement phase, neutron flux is reduced down to 50%.

The Miniature Neutron Source Reactor is a tank-in-pool research reactor that uses highly-enriched uranium as fuel, light water as coolant, and beryllium as reflector. During the operation of the reactor, the coolant temperature increases and leads to negative feedback in the reactor. Although the negative reactivity is amongst the advantages of this reactor and an inherent safety feature; it causes a decrease in both the additional available reactivity and the reactor operation time. Therefore, short operation time (about 2.5 hours) at nominal power is one of the main limitations of the MNSR reactor. The natural convection in the core, vessel and pool of the reactor was modeled through a CFD analysis and the effect of the cooling coil at the top of the reactor tank on increasing the reactor operation time was investigated. To reduce the computations the details were decreased by considering the reactor core as a porous medium and a heat source with a constant power of 30kw. The experimental values ​​and those ​​obtained from the numerical solution are in good agreement. Results show a steady upward slope in the temperature rise of the coolant in the absence of coil, and an about-2-hours rise of the temperature in its presence. After this 2-hours period, the increasing rate decreases and the temperature fluctuates in a certain range. Compared to the case without the cooling coil, the average temperature is reduced by 3 degrees and the reactor operation time is increased by 1.5 hours.

In this work, the Isfahan Miniature Neutron Source Reactor (MNSR) is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation (when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core) and also after 14 years of operation (total operation time of the reactor) is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time) is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

There are several methods to measure the reactor power. Since the power is directly related to the neutron flux in a reactor, the usual way to measure the power is to detect the neutron flux in the reactor core. The miniature neutron source reactor (MNSR) uses two fission chamber (FC) neutron detectors connected to computer and console control systems to determine the reactor power. These instruments are indicators of power, therefore the output of them has a significant role in reactor safety and needs to be calibrated. Importantly, the calibration measurements should be performed in a neutron-gamma environment very similar to the environment in which they will be used later. The aim of this study is to calibrate the fission chamber detectors in the core of the reactor by using the solid-state nuclear trace detector (SSNTD). The calibration factors of the FC detector related to the computer and control systems are equal to 1.02 and 0.88, respectively.

In this work, the relative neutron flux was determined experimentally using neutron activation analysis (NAA) method along the length of the dry channel (GUIDE TUBE) of the Isfahan miniature neutron source reactor (MNSR). This reactor was also simulated using the MCNP code, and the neutron flux distribution along the dry channel was calculated and the results were compared with the measured values. The results showed that the neutron flux distribution peak in the dry channel occurs at a point below the nearest point to the center of the reactor core. This is due to presence of the bottom beryllium reflector. In addition, the simulation program was used to determine the neutron energy spectrum in the dry channel and also in the inner and outer irradiation channels of the reactor. Furthermore, the neutron energy spectrum in an inner irradiation channel of the reactor was compared with the previous studies.

Neutron radiography is an unique, advanced and useful non-destructive test method in various industries and researches. Nuclear reactors are powerful and stable neutron sources for the neutron radiography system. In this research, the MNSR research reactor has been used as a neutron source for a neutron radiography system, and its neutron beam parameters have been evaluated. Also, using the direct conversion method and the gadolinium converter, the neutron image was obtained in its neutron beam and the image quality of the neutron was evaluated using the standard image quality standard ASTM-E545. The results of the measurements show that the neutron flux of the system in the maximum reactor power is equal to 3×105 n.cm-2.s-1 and the collimation ratio is 100. Image quality evaluation also showed that the obtained image has a V quality category based on the ASTM-E545 standard.

One of the important neutron sources for Boron Neutron Capture Therapy (BNCT) is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA). In this paper، Miniature Neutron Source Reactor (MNSR) as a neutron source for BNCT was investigated. For this purpose، we designed a Beam Shaping Assembly (BSA) for the reactor and the neutron transport from the core of the reactor to the output windows of BSA was simulated by MCNPX code. To optimize the BSA performance، two sets of parameters should be evaluated، in-air and in-phantom parameters. For evaluating in-phantom parameters، a Snyder head phantom was used and biological dose rate and dose-depth curve were calculated in brain normal and tumor tissues. Our calculations showed that the neutron flux of the MNSR reactor can be used for BNCT، and the designed BSA in optimum conditions had a good therapeutic characteristic for BNCT.

It is shown recently that the thermal neutron field of Isfahan MNSR with stable flux and dose equivalent rate can be used as a calibration field in dosimetry. The fundamental limit of this field is not the feasibility of using phantom for the irradiation of personal dosimeters. This issue leads to underestimating the personal dose-equivalents from the true values. The subject of this work is to correct the calibration curve of TLD-600 dosimeters irradiated without phantom. To do this, Monte Carlo simulations using Geant4 are carried out and absorbed doses in this dosimeter for irradiation with and without phantom are calculated. Then, the ratio of these doses as a correction factor applies to the responses measured without phantom, and then the corrected calibration curve is determined. As an analytical approach, the correction factor is considered as the ratio of the mean number of thermal neutron reflections between air and water (with phantom) and in the air alone (without phantom). Results obtained show that the correction factors determined by the simulation and analytical methods are 1.57 and 1.44, respectively, which agree well with the 8% difference. Finally, the response correction has led to changing the calibration factor of TLD-600 dosimeters from 0.0012 to 0.0008.