محمدامین امیرخانی دهکردی

Identification of the changes of aluminum structure due to neutron radiation is very important in the reactor fuel clad. In some cases, ion (usually proton beam) irradiation method is implemented for this investigation. The changes in the holes of Al are in the range of picometers by the radiation of energetic protons. In this research, positron annihilation lifetime spectroscopy, was used to investigate the changes of holes in Al-6061 and Al-303 due to the radiation of 2.2 MeV protons at different amounts of radiation. This method has the ability to evaluate the dimensions of the holes in the range of picometers and nanometers. For both investigated aluminum samples, the results show that the value of τave first decreased and then increased. It indicates the holes became smaller and decrease the vacancy defects in the initial stages of irradiation. Then, the holes became larger in the continuation of irradiation, which indicates increasingin the interstitial defects. These results were also confirmed by XRD test. It can be seen that the crystal size of the irradiated aluminum increased at the in the initial stages of irradiation and then decreased in the continuation of irradiation.

At present, Maximum Temperature Crystal Sensors (MTCS) are of great interest in the design and optimization of engines and turbines with high efficiency. These sensors are made in very small dimensions and can measure the maximum temperature in the desired location, for example inside the turbine. It is worth mentioning that a high fluence of fast neutrons is required to produce these sensors. The MTCS method is based on the high temperature phenomenon to restore changes made in the crystal lattice caused by neutron irradiation. Calculations of the maximum fast neutron flux for the diamond sample were performed using MCNP6 with the simulation of the core of the Tehran Research Reactor (TRR). Activity calculations were done using ORIGEN2, dose rate calculations through MCNP6, and temperature calculations with ANSYS software at different points of the sample. The highest fast neutron flux was calculated, and the activity of the sample decreased significantly after 40 days, with the dose rate at a distance of one meter reaching 1.4 µSv/h. Temperature calculations showed that the maximum temperature at the lowest speed of the channel was 402 K, which would not cause any thermal problems. Additionally, using the X-Ray diffraction (XRD) method, the lattice number and unit cell volume were calculated for the sample. The results obtained suggest that the diamond sample, from the perspective of neutronics, dosimetry, thermohydraulics, and changes in crystal lattice structure, can be recommended as a crystal sensor in the Tehran Research Reactor.

In recent years, one of the methods that is considered in designing and optimization of the engines and turbines with high efficiency is maximum temperature crystal sensors (MTCS). These sensors are usually very small in size and made of materials such as 3C-SiC, Diamond, Graphite, etc. They can be placed in the desired location inside the turbine or engine and measure the maximum temperature created. These sensors could be obtained a very accurate spatial distribution of the created temperature due to the small dimensions and the very high accuracy of the measurement. For producing these sensors, a very high fluence of fast neutrons is required. Therefore, the number of the core 18 of the Tehran research reactor (TRR) is simulated and the calculations of the maximum flux of fast neutrons were carried for the 3C-SiC sample. The simulation results show that the channel F4 will provide the highest flux of fast neutrons for this purpose. Also, activity calculations were done using ORIGEN code and dose rate through MCNP6 code. The results show that the amount of activity of the sample has decreased by 65% after 4 months. Therefore, the total dose rate has also reached a value lower than the criteria of 10 µSv/h. Next, the 3C-SiC sample was irradiated at 5 MW power in the TRR, then the crystal size and the amount of strain before and after irradiation were calculated. Finally, according to the results of this article, it can be said that the 3C-SiC sample will be suitable as a MTCS.

Radiation damage calculation is very important for estimating the lifetime of nuclear power plant instruments. The purpose of the present study is to evaluate the sensitivity of the damage values to the neutron group energy, where the neutron flux rate is maximum. In the present study, we propose three neutron group energy (WIMS, CINDER, and OPENMC) to evaluate the sensitivity of radiation damage calculations. The obtained result from SPECTER code shows that the differences between these results and the standard values (ASTM E-693) for three group energy WIMS, CINDER and OPENMC are respectively 0.25E-04, 0.75E-04, and 1.74E-04 in the ¼ diameter of reactor pressure vessel (RPV). Therefore, the WIMS energy group is the most accurate spectrum compared to other spectrums in this region. Also, these differences are respectively 0.43E-04, 0.52E-04, and 1.86E-04 in the ¾ diameter of RPV. Moreover, the WIMS group energy is the most accurate spectrum in this region. This result shows that increasing the number of neutron energy group cause to reduce the difference between calculation of damage and its standard values. According to the results of calculated damage that induced by WIMS group energy in ¼ diameter of RPV, the values of yield strength at 1, 5, 10, 15, 20, 25, 30, 35 and 40 years are calculated by SPECOMP and SPECTER codes.

Change in the position of atoms in the material’s crystalline lattice is caused by the radiation. Changes in the macroscopic properties of the material will be created by the change in structure. Knowing the amount of variation of material properties is necessary for materials in high-radiation environments. In this paper, the effect of 2.5 MeV proton irradiation on TSX grade graphite structure is investigated. This graphite is used as a reflector in the Tehran research reactor. The radiation has been carried out with the proton for 276 minutes. Microhardness test, Raman test, and SEM images were used in this study. The results show an increase in the microhardness of the material due to the radiation. Increased vacancy and interstitial clusters have also been observed in the Raman spectrum. The creation of cavities on the surface of irradiated graphite is observed in SEM images. The results can be verified by using the point defect model.

The displacement of the atoms from their lattice sites is one of the results of neutron irradiation on materials. The primary knocked-on atoms spectrum and their angular distribution should be calculated for consideration of neutron radiation damage calculation. The AMTRACK program has been developed to calculate this information. This program extracts and analyzes information about the primary knocked-on atoms by using Ptrac output of the MCNPX code. In this study, Stainless Steel 316, one of the most important alloys in the reactor pressure vessel, is investigated. The material is irradiated by single energy neutrons of 1keV to 10MeV, the fraction of produced PKAs, their average energy, their maximum energy, and radiation damage value are calculated. The calculations are performed by using the neutron spectrum of the Bushehr reactor. Using this method, the amount of damage in the Bushehr reactor pressure vessel (for Stainless Steel 316) is equal to 7.2×10-22 (dpa/ fluence).

Safety is one of the most critical issues in any test. The flexural test is a significant one to obtain material specifications. In this paper, a standard sample dose rate was calculated for the graphite flexural test in Tehran research reactor. Using different standards, the dose was computed at different distances and two periods of radiation, 15-40 days. Our calculations were performed with the ORIGEN and MCNPX codes. The impurity amounts of the material were measured by two methods of X-ray Fluorescence (XRF) and Inductively Coupled Plasma, Atomic Emission Spectroscopy (ICP-AES). The obtained results show that the Claiborne & Trubey standard is more stringent than other standards, and it has less than 2 μSv/h dose with the observance of 100cm distance and 20 days after irradiation. Moreover, calculations of the transfer chamber indicate that there is no limitation for the transport of the samples. Finally, it can be concluded that the considered method of this research can be used also for other materials at the reactor core.

Spallation process is the most important neutron generation method in industry, medicine, etc. This process in the subcritical reactor core is also another technique. In this research, we study the neutronic behavior of the spallation targets consisting of U-238 and Th-232 materials, by MCNPX code. The parameters under study comprise the spallation neutron yield, deposition energy, target geometry; angular spectrum of the neutron output, gas rate and residual mass spectrum. The results show that geometry has the greatest impact on the neutron output spectrum, but not on the residual mass spectrum. Numbers of neutrons per energy unit are stable at higher energies of 1 GeV, then the changes in neutron generation rate are reduced. Furthermore, hydrogen which is the principal factor in swelling of spallation target, consists of about %88 of gas production.