جستجوی مقالات مرتبط با کلیدواژه « رآکتور mnsr » در نشریات گروه « فیزیک »

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.

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.

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 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.

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.