نشریه سنجش و ایمنی پرتو
سال هشتم شماره 4 (پیاپی 33، بهار 1399)

Strontium Tetraborate doped with Lead SrB4O7:Pb powder was prepared by solid state method. The formation of Strontium Tetraborate compound was analyzed by X-ray diffraction (XRD) and Scanning Electron Microscopy (SEM). Thermoluminescence Glow Curve of gamma irradiated sample was investigated in order to use as a thermoluminescent dosimeter. Because of excellent mechanical properties, non-hygroscopicity and sensitivity to gamma radiation, SrB4O7:Pb can be used as a personal dosimeter.

In this study, calibration process was carried out for deigned new CR-39 nuclear track detector for protons, neutrons and alpha particles separately under the same etching condition. In order to aim this purpose, americium-beryllium standard source (241Am-Be) and Plexiglas phantom for neutron irradiation, brass collimators and americium standard source (241Am) for alpha irradiation and the accelerator and Van de Graff accelerator for proton irradiation were employed. Sodium hydroxide solution 6.25N at the temperature of 85 °C was used for CR-39 etching. Considering obtained results, different detection shields were designed to distinguish between fast neutron particles, thermal neutrons, albedo neutrons, protons and alpha in mixed radiation fields. Moreover, both the contribution of each particle and the ability of the designed detector to discriminate energy of charged particles were found.

Recently, proton therapy is used as one of the effective methods for treating various types of cancer in clinical treatment. An appropriate formalism to obtain relative biological effectiveness values for treatment planning studies is needed in this hadrontherapy technique. Hereby, the quantity of biological dose, instead of using the physical doses, is introduced to evaluate the biological effect of radiation in tissue. In proton therapy, a constant RBE of 1.1 is usually applied clinically as recommended by ICRU. However, different studies including data from irradiation experiments propose that a variable RBE factor is more appropriate. The primary objective of this project was to calculate and compare the dose-to-patient results of constant RBE versus variable RBE calculated by different models for hadron therapy. The irradiation experiments show the RBE parameter depends on different parameters such as deposited dose, LET, radiation energy and tissue sensitivity. In the present study, Wilkens, Wedenberg, Carabe and McNamara models were used to calculate RBE and biological dose in different models by using GATE Monte Carlo code. The results of the different models were compared with constant RBE of 1.1. On the other hand, bringing further simulations closer to the real situation, proton beam modulation has been used to create Spread-Out Bragg Peak (SOBP) region, and different models of RBE have been used to calculate the biological doses. Results show that the four different models predicted an equal or lower dose in the proton beam-entrance region compared to predictions for constant RBE, while greatly exceeding the predicted constant RBE dose in the distal part of the region target.

In the last decade, several studies have concluded that elevated concentrations of radon gas in soil or groundwater could be the sign of an imminent earthquake. It is believed that in case of sudden movement of rocks before earthquake, radon is released from the depth of ground and reached the surface. In this way, by measuring Radon concentration change, earthquake prediction is possible. In this study, for the purpose of the feasibility of using microstrip gas detector in monitoring Radon change, initially for the verification of the code, the range of radon alpha particles and its progeny in the air has been obtained in the standard condition, which is consistent with the experimental results. Then detector geometry has been simulated by MCNPX code and after examining the average of detector response to decay, the output code is calibrated considering the information about the radon changes in Jowshan hot spring. The results have shown that the detector response is comparable to the experimental data and can be used as an earthquake pre-indicator.

Neutron Generators (NG) are used as a neutron source for different applications. During recent years, major efforts are underway to develop a high yield compact NG. In this way, radiation protection aspects need to be considered during the operation of these high yield NGs. In this paper the neutron effective dose of a NG operator has been calculated using MCNPX Monte Carlo code. The results show that the neutron dose around a NG tube is highly dependent on the angle, distance and the type of NG (DD or DT). The dose due to DT source is 500 time higher than the DD source. Increasing the distance from 1 m to 5 m will decreased the dose up to 20 times. Increasing the distance is the effective way to reduce the dose rate but in a laboratory which there is not enough space, an appropriated neutron shield should be considered.The shields had designed in 6 different materials (ALF3, Borated-Polyethylene, concrete 806, Paraffin, Polyethylene,non-borated, Solid boric acid) and in thicknesses of 10, 20, 30, 40, 50 and 60 cm. The gamma flux, gamma effective dose, thermal neutron flux, epithermal neutron flux, fast neutron flux, total flux of neutron, and neutron effective dose components had calculated in a phantom spherical hypothetical shape. The results showed that Borated-Polyethylene shields for both of DD and DT source in thickness of 60 cm had the minimum neutron effective dose and also this shield in 60-cm thickness had lower gamma Dose than the other shields in both of the DD and DT source.

Unlike proton therapy, conventional radiation therapy directs X-rays not only at the tumor but also unavoidably at nearby healthy tissue. Protons deliver radiation to tumor tissue while the healthy structures will be spared during proton therapy. When protons travel through matter, secondary particles like neutrons and photons are produced. It is believed that the secondary dose can lead to secondary cancer. It is not generally possible to measure directly the absorbed dose in a Human body. Therefore, Monte Carlo simulations and phantoms are useful for estimating the absorbed dose in organs. In this study, the MCNPX and Analytical phantom of the human body, ORNL-female phantom were used for breast proton therapy. We considered a characteristic tumor with the same composition of breast tissue. Measurement was performed with the energy range of 60-70 MeV proton beams accelerated. Secondary productions generated in the body of the patient can affect the organs surrounding the target region, so the equivalent dose of these secondary radiations was calculated in vital organs including Right-Lung and Left-Lung. For all energies, the equivalent dose of photons in vital organs is lower than neutrons. Our Monte Carlo results show that the equivalent dose of the Neutrons and Photons in the left lung is 0.1051 mSV and 0.0178 mSV in the right lung per 1mSV of 70 MeV energy incident proton beam. It is shown that for high-energy proton beams, most of the absorbed dose by organs is due to secondary neutrons but those are low enough to be neglected. Breast proton therapy provides satisfactory target coverage and enhances normal tissue sparing that can limit unnecessary doses delivered to the lung, though there has been a significant development in Radiation therapy technology, there still remain concerns on treatment related to long-term side effects. This problem is more pronounced in pediatric cases.

One of the most important groups that can be exposed to nuclear radiation is pregnant women and their fetus. This group of people can be technicians at nuclear centers located in the patients undergoing treatment or nuclear medicine diagnostic centers. The purpose of the paper is to determine the rate of doses received by female technicians who are indirectly in the vicinity of nuclear radiation. In order to verify the results of computer simulation, a comparison was made with experimental results from the nuclear medicine department of Imam Hossein Hospital in Shahroud. The statistical analysis of the results showed that there was a significant difference between the simulation results and there are no experimental results.

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.

Technetium is one of the most important radioisotopes recognized in medicine which is obtained through ‎the decay of molybdenum 99.The half-life of this radioisotope is 6 hours and it is capable of 140 kev gamma ray emission. Due to its short half-life, this radioisotope must be produced at the site of consumption, so that the shortest possible time interval between production and consumption can be achieved. The aim of the present study was the neutronic ‎evaluation of the amount ‎of 99Mo-99TC production through nuclear fission in the Tehran research reactor. The primary core ‎of the reactor with the code MCNP X 2.6 was simulated and neutron flux was assessed in 6 locations. ‎Then, the fuel plate was placed at the site with the highest neutron flux. In different phases, changes ‎were made in the enrichment of fuel. Moreover, the amount of molybdenum produced in each phase was ‎calculated. Based on the results, the amount of‎ molybdenum produced, and thus,‎ the amount of ‎technetium were the highest in the 20% enriched fuel.

In this study, 34 soil samples were collected from the east of the Shazand power plant to the  Arak city, with a length of 25 km. The specific activities of radionuclides of 226Ra, 232Th, 40K, and 137Cs were determined using gamma-ray spectrometry method. Gamma-ray spectra registered to employ high purity germanium (HPGe) detector with 30% relative efficiency. Specific activities of corresponding radionuclides in the soil samples varied from 18.92±3.82 to 43.11±5.38, from 25.31±4.23 to 54.27±7.30, from 230.17±19.93 to 728.25±36.06 and from <1.49 to 35.77±1.76 in Bq/kg, respectively. Alpha and gamma indices were calculated. Gamma index with average value as 0.86 in samples varied from 0.61 to 1.16.  Alpha index obtained in a range of 0.09 to 0.22 with average as 0.13. Radiological map of radionuclides distributions using SURFUR15 software was derived.

Thorium fuels have attracted researcher’s attention due to the reduction of uranium deposits in the world, more abundances than uranium, and the ability to be breeding in thermal and fast reactors. This study aimed to investigate the radiation safety and the amount of necessary shielding for the transportation of spent thorium fuel in the Tehran Research Reactor. The ORIGEN and MCNPX computational codes were used to calculate the gamma spectra of the spent fuel and the fuel dose rates loaded within the lead cask. The results of the calculations showed that compared to uranium fuels used in the Tehran Research Reactor, more cooling times are needed before the thorium spent fuel transport by the lead cask. In addition, the thickness of the cask required for the transport of the spent thorium fuel is greater than uranium fuel.

Measurement of ionizing radiation in various fields, such as environmental safety, industrial detection processes, radiation protection and medical is very important. Radiation dosimetry plays an important role in determining the amount of energy absorbed and the effects of radiation. Recent optical fiber sensors have been shown to be radiation dosimeters. Our goal here is to investigate the effect of ionizing radiation on fiber optics. Thus, in this paper, the effect of electron beam radiation on the loss of light passing through the fiber optics after the end of radiation in the wavelength range of 1300-1550 nm is investigated. Optical fibers with doses of of 22 and 47 kGy were irradiated. All measurements were carried out at a temperature of 25 ± 2 ° C. The results show that the light dissipation of the fiber optically increases with the radiation dose and decreases with the passage of time after the end of radiation.

In this paper, digital signal processing for precise gamma ray spectroscopy is presented. The basis of this system is a 14bit waveform digitizer which samples the output of pre-amplifier signals directly. The advantages and limitations of the digital spectrometer compared with analog spectrometer has been tested and analyzed from precise gamma ray spectroscopy point of view. The results shows that the digital pulse processing provides better energy resolution, longer stability in time, better peak to Compton ratio, higher throughput and effective elimination of ballistic deficit.

One of the main factors for satellite design is simulating of total ionizing dose due to space ionizing rays in devices used in space. By measurement of induced dose based on available data in different altitudes, expenses of designing, satellite weight and amount of needed fuel will be reduced. Optimum design of satellite for protecting satellite against ionizing radiation has considerable effect on reduction of manufacturing expenses. TID effect should be calculated for each mission separately. The main parameters for starting simulation are perigee and apogee altitudes, inclination, launch date and space environment condition. In this article, the induced dose for missions with equal apogee and perigee points and the most dangerous altitudes for launching satellites have been simulated. To gain this goal, simulations have been performed in different inclinations from 500km to 58000km. by plotting result of all these simulations for launching satellite in minimum solar activity, a dose map has been achieved for different altitudes and inclinations. Absorbed dose in this dose map is sum of absorbed dose of all radiations in space environment. Simulation results illustrate that the maximum dose are related to low inclinations in altitudes between 11000 and 12000. The result of calculated and simulated results have been confirmed and evaluated by ECSS documents. These results have been simulated by using the last version of OMERE radiation package.

Gamma-ray detection has an important role in the enhancement the nuclear safety and provides a proper environment for applications of nuclear radiation. To reduce the risk of exposure, aerial gamma survey is commonly used as an advantage of the distance between the detection system and the radiation sources. One of the most important issues in aerial gamma survey is the detection noise. Various methods being proposed to reduce the noise of the gamma detectors, among which, in this paper, the utilization of Cyclo-stationary properties is proposed, because of its capability in detecting weakened gamma rays with low rate counts from far sources. To increase the accuracy of the results of gamma detection and reduce errors due to physical and flight constraints, we compared other time-series processing and spectral estimation methods. The most important problem with such methods is the high computational complexity, which makes them difficult to use. In this paper, we present the aerial gamma detection noise reduction methods based on the Cyclo-stationary properties in extended Kalman filters. The Kalman estimates real-time variations in the counts of radionuclides using data integration based on a dynamic model. The results show that the extended Kalman is superior to other filters due to its nonlinear distortion reduction feature. The focus of the paper is on the modeling, matching and computational aspects of applying the Kalman filter on real data obtained from aerial gamma survey. The covariance and the required computational time have been evaluated using the power spectral density estimation, Multi-taper spectral estimation, and the extended Kalman methods. The results indicate that the extended Kalman method increases the converging speed in addition to empowering the detector against the nonlinear noise and disturbances.

During radiotherapy by any radiation, it is always essential to stop absorbing the excess dose by a tissue. To better treat cancerous tissues and to make more precise irradiation for a cancerous tumor, there needs the accurate irradiation time to be estimated. First, the constituent materials of any of the existing organs in abdominal tissue are extracted and defined in the MCNPX code. Then, every organ in the abdominal tissue is voxelized by MATLAB software. Each of the voxels is defined based on Hounsfield unit of pixels in DICOM images. Then, the voxels are assigned to the related tissue which is comprised of its constituent materials, and they are filled up with them. Then, the liver tissue is segmented among the abdomen region from other tissues. Then, the geometry of the segmented liver tissue is generated as input data for MCNPX code, and the absorbed dose is computed. After obtaining the values of absorbed doses in liver tissue per each of incident fast neutron energies, the required irradiation time is obtained in second by making an appropriate proportion of absorbed dose and activity. This precise required irradiation time is obtained by an advanced software application (designed in this research using Delphi 7 programming language) through establishing a relationship between the absorbed dose and activity based upon the energy of clinical fast neutron source. The required irradiation time is calculated to reach the desired dose for each patient during fast neutron radiation therapy for similar liver tissues accordingly.

The purpose of this study is to estimate internal dose to human body due to administration of I-131 using new biokinetic model of iodine. In this respect, we used international commission on radiological protection (ICRP) adult reference phantoms. Monte Carlo calculations were performed using MCNPX 2.6 code. The average dose to target organ per disintegration inside source organ, namely S factor, was estimated using F6 and *F8 tallies, and then compared to those from previous studies. The results show that the difference between S factors is on average less than 8%. Furthermore, the variations caused by different 131I decay spectrums were investigated. Finally, the absorbed and the effective doses were obtained for adult female and male ICRP reference phantoms considering two biokinetic models. To conclude, using different tallies and decay spectrums, as well as, various biokinetic models could affect considerably on the amount of estimated internal dose.

Determination of the relative biological effectiveness (RBE) of Auger electrons is a challenging task in radiobiology. In this study, we have estimated the RBE of internal conversion (IC) and Auger electrons released during Gadolinium neutron capture reaction (GNCR) by means of biological weighting functions (BWFs) with microdosimetric approach. Regarding the different distribution of Gadolinium (Gd) relative to the DNA as a target, the microdosimetric parameters of the Gd electrons were calculated using the Geant4 Monte Carlo toolkit and ROOT software. Assuming Gd infiltration into the cells and uniform distribution inside the Cell, the lineal energy distribution of Gd electrons in DNA was used instead of the lineal energy distribution of external radiations in the micrometer-diameter targets, which has been conventionally used in the mentioned methods.The results show that the calculated RBE values of Gd electrons using BWFs (2.68) for the case where Gd distributed at the center of the DNA are approximately equivalent to the RBE value of the therapeutic neutrons, which were measured in the literatures with the same biological conditions. According to the results, although the changes of the RBE of Gd electrons to the different distribution of Gd relative to the DNA are small, the amount of biological dose of the Gd electrons in the DNA is strongly dependent on the Gd distribution. In the case where Gd distributed at the center of DNA, the mean biological dose of Gd electrons in DNA during one GNCR (227.8 kGy.Eq) is large enough for occurring double-strand breaks (DSB) of the DNA. If we have accurate information about the spatial distribution of Gd or Auger-electron emitters inside the cell, by comparing to the results obtained in this study, we can have a better estimation of the RBE of the Gd electrons or in general Auger electrons.

According to International Atomic Energy Agency (IAEA) recommendations, the calibration of ionization chambers used for radiotherapy dosimetry follows the substitution method which demands the use of a reference ionization chamber. This work introduces the Design and fabrication of cylindrical ionization chamber for dosimetry to be used by standard dosimetry laboratories. The result of the quality control tests  in Secondary Standard Dosimetry Laboratory (SSDL) of the Atomic Energy Organization Which was carried out in accordance with the standards and limits set by the International Atomic Energy Agency, are evidence of the accuracy of the claim that this chamber can be used as a reference dosimeter in standard dosimetric laboratories.The low polarity effect and leakage current, high ion collection efficiency and stability, linear response of the detector to the dose and the rate of dosing are some of this instrument characteristic through optimum design of guard ring and collecting electrode and also the proper choice of materials in fabrication.

Radiotherapy using various beams is one of the methods for treating cancer, Hadrons  used   to  treat cancers  that  are  near critical organs. The most important part of the cell that is damage by ionizing radiation is DNA. In this study, damages induced in the  genetic material of  living cells (DNA) defined by  the  atomic model from the  protein data bank (PDB) have been studied by  radiation of monoenergetic protons and its secondary particles using  Geant4 code. The total SSB yield is independent of energy of incident particle. The total DSB yield is increase with increasing of  incident particle LET. The ratio of total inelastic events to absorbed dose for primary protons as well as its  secondary particles is independent of energy. The contribution of  secondary particles in the formation  of  single-strand  breaks and double-strand breaks have been calculated. The DSB yield generated by secondary particles is increase with decreasing the energy of the incident particle and the contribution of  secondary particles to the generate of double-strand break for energies less than 5 Mev  is greater than their contribution to the creation of single-strande breaks. The ratio of double-strand breaks to single- strand breaks have been reduced by increasing  the energy of   the radiation particle.

Given the extensive use of common mammography tests for screening and diagnosis of breast cancer, there are concerns over the increased dose absorbed by the patient due to the sensitivity of the breast tissue. Thus, knowing the Mean Glandular Dose (MGD) before radiation to the patient through its estimation can be helpful. For this reason, the MultiLayer Perceptron (MLP) neural network model was trained with Levenberg-Marquardt (LM) backpropagation training algorithm and the Entrance Surface Air Kerma (ESAK) was estimated. After running the program, it was found that 35 neurons is the most optimal value, offering a regression coefficient of 95.7%, where the Mean Squared Error (MSE) for all data was 0.437 mGy, accounting for 4.8% of the range of output changes, representing a prediction with 95.2% accuracy in the present research. In comparison with the Monte-Carlo simulation method, it enjoys a desirable accuracy.

The radioactive contamination measurement is essential in nuclear facilities. Monitoring of low level alpha and beta contamination is possible using light detectors high efficiency. Due to the fact that many detectors canchr('39')t simultaneously discriminate alpha, beta and gamma rays; among existing methods, phoswich detectors would be appropriate option to measure different radioactive contaminations. By this method, it is possible to study simultaneously the alpha/beta/gamma contamination sources with a single detector, which consists of three layers of different scintillators. Here, our design using ZnS(Ag), BC400 and NaI(Tl), with thicknesses of 0.0025, 1 and 3 cm, is sensitive to interaction with alpha, beta and gamma particles, respectively. The absolute efficiency for each layer was determined by the trade-off between the efficiency and background counts due to interference radiation in each layer. Thus, at an energy of 0.1 MeV, the maximum absolute efficiency for beta particles was 24% and 1-2 MeV at maximum absolute yield of gamma rays with an approximate value of 16%.

High Purity Germanium detectors (HPGe) are subdivisions of semiconductor detectors which are widely used in nuclear technology from space industry to nuclear medicine, due to their high resolution, low dead time, unlimited size and compatibility with a variety of environments. The( absolute and intrinsic) efficiency of the HPGe detector, which depends on the geometry of the source-detector system and also on the energy of gamma ray, is an important factor in determining the activity of the radioactive sources. A simpler way to determine the activity of radiation sources is to use the virtual point intrinsic efficiency which is independent of source-detector distance. The virtual point is a point within the detector that all interactions are assumed to take place in that point. The goal of this study is to determine the location of the virtual point of the HPGe detector of GMX 40P4-76 model and determine the virtual point intrinsic efficiency for this detector at various energies of gamma rays. To do this, we first experimentally obtained the spectrum of gamma rays emitted from a Europium source at different distances from the detector using the HPGe spectroscopy system. Similarly, using the Monte Carlo simulation method (MCNPX code) we simulated the spectroscopy system. In this study, in the first step, the virtual point distances from detector top were determined in two experimental and simulation methods for gamma ray energies from 121 keV  up to 1408 keV, the results of which were in good agreements. In the second step, using the virtual point distances, we calculated the virtual point intrinsic efficiencies at various gamma ray energies for the HPGe detector, and it was shown that this efficiency for a given energy  is almost constant with small deviations for various source-detector distances and it is therefore easier to calculate the activity of unknown sources using this efficiency, because the distance dependency is gone and we do not need to perform  measurements at different distances.

In this research, the applicability of several metal hydrides as neutron moderator and shielding for D-D fusion sources has been investigated by MCNPX code. The results have been investigated in three steps to find the materials with lower thermal, fast and total neutron fluxes than conventional shielding materials. The results show relative advantages of LaNi5H6, VH, TiH2, TaH, Mg (BH4)2, YH2, NbH and LaH3 metal hydrides in reducing the energy of D-D neutron sources, so that utilization of these hydride compounds as a part of the neutron moderator or shielding will be very effective in reducing the volume of nuclear facilities.

Nowadays, due to the high costs and large dimensions of the conventional proton accelerators, other optimal methods for producing the proton beam have been studied. Using of Laser-driven proton accelerators is one of the important and new methods. In laser-driven ion acceleration, a highly ultra-intense laser pulse interacts with solid density targets and will create a plasma media that will accelerate ions and produce protons. Currently there are projects in this in field such as ELIMED, PMRC, DROT, HZDR and …. Due to the large angular dispersion of protons, reducing their dispersion and collimating them for transmission is very important. In this research, using the GEANT4 toolkit, the exact solenoid and its magnetics field for HZDR beamline have been simulated. The effect of solenoid on protons with a primary divergence of 5 degrees was investigated. The results show that consideration of the details of the solenoid has a direct effect on the calculation of the proton profiles and is necessary. The maximum of the solenoid magnetic field was calculated to ~20 Tesla to fit the proton selector system at a distance of one meter from the source.

Radiation detection is essential for determining of radiation dose. Depend on the detector and dosimetry method, detection process is performed in different levels. Pulse counting is the first level of detection. Typically, the output of a radiation detector for determining value of the radiation dose cannot be used directly. Through changing the response function or the readout detector, is trying to create relevance and proportionality between detector readout and value of the dose equivalent. For this purpose, there are various software methods and hardware methods are obtained dose equivalent values. Software methods such as convolution, Deconvolution, etc. and hardware methods such as additional of layers of modulators, etc have been used. Selective data sampling of multi-channel energy is a software method which has been studied in this paper. The main purpose of this method, is obtaining the response of photon dosimetry at different energies. The response function of NaI(TL) 3”×3” detector for 0.411 MeV to 3 MeV gamma-rays has been simulated by using MCNPX code, and for calculating dose, the calibration coefficients were determined. Finally, the response energy of the detector has been drawn for photon dosimetry of the channels with multiple energy range. Results show that presented dosimety method has high level precision.

The giant magneto impedance (GMI) effect is a large variation in the electrical impedance of a magnetic conductor when subjected to a static magnetic field. The sensitivity to the direction (AGMI) and magnitude of applied  magnetic field    and also linearity levels   of this  effect are three important parameters in magnetic sensors application. A suitable annealing procedure can be used to achieve an optimum response of GMI by reducing the quenched-in internal stresses and induces transverse magnetic anisotropy.  In this research the effect of irradiation of gamma rays from two gamma sources (60Co, 137Cs) on the asymmetry of GMI  of Co68.15Fe4.35Si12.5B15  amorphous alloy is studied. The optimum asymmetry of studied sample obtained after 30 (24) hours irradiation by photons emitted by 60Co (137Cs) source. The X-ray diffraction pattern shows a transition from amorphous phase to a crystalline phase in  samples. Results  show that induced changes by irradiation with 137Cs source leads to a better sensitivity  to  direction of applied magnetic field especially in case of irradiated sample for 24 hours. This sample can be considered a good magnetic sensor candidate.

Brachytherapy, is a method to treat prostate cancer in which a radiation source is placed inside or next to the cancer affected tissues. Purpose of this article is to determine the uncertainty level of received dose in prostate tissue due to the relocation of the placed radiation sources and inflation in prostate after positioning the radiation seeds. To simulate the model 6711 Amersham (activity level equal to 0.5 mCi) iodine-125 brachytherapy source, we used MCNPX 2.6 code and TG-43U1 protocol, and we used ORNL phantom. Treatment plan included 76 sources consisting iodine-125 radioactive, once placed in the form of seeds and once applied in the form of beads in three initial volume: 30.02, 38.01 and 52.01 cm3 and three different time steps were placed in prostate phantom. Three-time steps were investigated, including the moment after placing, the zero days and thirty days after placing. In the first step, the obtained dose of radiation in three different prostate volumes, for both seed distribution and bead distribution in healthy and cancerous tissue, was calculated. In the second step, position of the sources was relocated from the assumption, and causes a change in dose distribution. Considering the movements in 3 directions of left-to-right (1.8 mm), intra-exterior (2.1 mm), up-down (3.4 mm) after implantation the equivalent dose was calculated. The maximum uncertainty of received dose in a prostate with seed implanted sources and the volume of 52.01 cm3 and it was reported as +20%. At the third step, in addition to the change in the location of sources, inflation rate of 12% was observed in comparison with the primary volume at the first step of treatment process which resulted in a smaller dose rather than the previous stage.In this treatment plan, relocating of the sources and 12% inflation of the tumor were analyzed. Uncertainty of source relocating was calculated as ±20% in zero-day of positioning. Also, 12% inflation of the tumor and source relocation after 30 days, in a prostate with the volume of 58.25 cm3 curing by point sources, shown 21.04% reduction in received dose for healthy and cancer tissues since the moment of placing. The maximum error of MCNPX code was calculated 0.03%.

Nowadays, in order to improve the accuracy of treatment in radiation therapy, there are many attempts to use magnetic resonance imaging (MRI) due to the advantages of excellent soft tissue contrast and ultra-fast pulse sequences. On the other hand, carbon-ion radiation therapy is developing rapidly due to the benefits of greater relative biological effectiveness (RBE) and the application in the treatment of some low linear energy transfer (LET) radiation-resistant tumors. The idea of using MRI guidance in treating carbon-ion, presents challenges including the dose perturbation in the patientchr('39')s body. For this purpose, in this study, using a Monte Carlo simulation, a rectangular phantom was modeled with various tissue layers simulating chest geometry of a patient with lung tumor. For the first time in this study, three-dimensional dose perturbation of a 220 MeV/nucleon realistic carbon-ion beam in the presence of two medium (1.5 Tesla) and high (3 Tesla) magnetic fields applied to the simulated tissue phantom, was compared with the non-field condition. Also, the distribution of the three-dimensional dose equivalent in the simulated heterogeneous phantom was calculated in the presence of a 1.5 Tesla (T) magnetic field. At the Bragg depth, no longitudinal displacement was observed for the centers of the dose and dose equivalent profiles affected by a 1.5 T field. The longitudinal displacement of the total dose profile in a 3 T field was calculated to be 1.1 mm. Furthermore, the amount of lateral deflection of the center of the dose and dose equivalent profiles in a 1.5 T field was equal to 1.7 mm, and the amount of lateral displacement of the center of the dose profile in a 3 T field, was calculated to be 3.3 mm. The results indicate that the dose perturbation are remarkable in the accuracy expected from carbon-ion radiotherapy guiding by MRI.

SPECT-CT is an imaging method for diagnosing diseases and studying physiology of the body. In this type of imaging, due to the use of radiopharmaceutical for imaging (SPECT) and X-rays (CT), a relatively high dose of radiation is introduced into the patient. If the patient is pregnant, considering that the developing fetus is sensitive to any radiation in the uterus throughout the entire period before birth, there may be harmful and irreversible effects on the fetus; therefore protecting the pregnant woman and the fetus against radiation is of great importance. The purpose of this study is to evaluate and calculate the fetus absorption dose in cardiac imaging using the SPECT-CT machine in stress and resting conditions of the cardiac by injecting 20 mCi of radiopharmaceutical of 99mTc-MIBI.  In order to evaluate the fetus absorption dose, the Monte Carlo method, using the MCNPX code and the ORNL phantom, was applied to simulate the fetus in three, six and nine months of pregnancy, and the radiation absorbed doses of fetus absorption were calculated in two condition of stress and rest.The absorbed dose of fetus in stressed state in three, six and nine months of gestational age are 1.26 × 10-2, 2.9 × 10-3, 2.83 × 10-4  mGy respectively and The fetus absorption dose in rest in three, six and nine months of gestational age are  2.52 × 10-2, 5.79 × 10-3, 5.68 × 10-4 mGy respectively. The results of this study indicate that the highest fetus absorption dose is related to cardiac resting imaging at the third month of the fetus. The fetus absorption dose has been calculated to be less than 0.05 Gy for all maintenance periods, which, according to the NCRP 128 and ICRP 84, do not cause a serious risk to the fetus. However, special immunological advices should also be given to pregnant women to reduce the absorption of fetus, especially in the early months of pregnancy.

In this paper, radiation effects in protection layers of logical cell of the digital gate in the FPGA for electron and proton rays was simulated Using the FLUKA Code. by using of the Monte Carlo simulation, the electron and proton transport into the logical cell of the digital gate in the FPGA will be studied. In this simulation, the maximum energy of the electrons and protons at the entrance of logic cell have been chosen between 30 to 50 MeV and the degradation effects of these rays on semiconductor material, their effects on five different layer configurations such as layer of aluminum, silicon, silicon dioxide, boron dioxide and boron will be studied. The simulation result shows that using the multi-layer of silicon dioxide relatives to the other cases leads to reduction of degradation effects on semi-conductor sensitive volume.

The origins of radon gas are the amount of uranium decay in the soil that is released into space and through the interior gaps of  buildings. Radon can enter the respiratory system and cause radiation hazards. Recently research has identified Radon as the second leading cause of lung cancer after cigarette smoking. The indoor concentration of Radon progeny is mentioned as an important issue regard to the point of view of health. Therefore, a more accurate measurement of the concentration of radon gas in order to assess the exposure of people is of great importance. The purpose of this research  is to measure radon gas with two contemporary and retrospective methods in Ramsar homes, which is known as a high level background radiation area, as well as comparison with previous measurements. In order to measure the Concentration of contemporary radon by  passive method using a radon chamber and solid state nuclear track detectors, a retrospective of the activity of polonium 210 implanted in glass objects was used to measure the concentration of radon gas. The solid state nuclear track detectors used in this research are polycarbonate, which is the first time used in this world for this type of polymer for the retrospective method. The average concentration of measured for contemperory radon is about 416 Bq /m3 and the resrospective radon is about 1299 Bq /m3. According to the comparison of the results  with the US Environmental Protection Agency standards, it was found that 45% of the radon gas concentration is above 148 becquerel per cubic meter, which is indispensable for reducing radon gas in these buildings. At the same time, the difference in the concentration of rare earth radiant gas with contemporary radon gas can be related to the change in the ventilation rate and the change in the concentration of aerosols and the ratio of the surface to the volume of the room.

The most important part of neutron therapy treatment (NCT1) is to achieve a beam of neutrons with suitable energy and intensity, as well as the least pollution and damage. In this study, in order to correct the neutron spectrum from D-D fusion and its use in neutron therapy, a set of different materials which are called the Beam Shaping Assembly (BSA) was placed in the direction of energy 2.45 MeV neutrons output in such a way that the neutron beam intensity and energy output were appropriate for treatment. Moreover, the total yield coefficient 109 was considered. In this article IS set for a plasma focus device was designed and optimized by MCNPX2.6 code. The results of this study indicated that, according to BSA3 materials, epithermal neutrons were the best choice for energy range.

Recently, a new Neutron Radiography (NR) beam line has been designed, constricted, installed and tested based on the use of E-beam tube of Tehran Research Reactor (TRR). Initial tests have been shown that the system can be used for different samples and purposes such as nuclear plates and rods fuels.  For this end the system need a suitable irradiation room which should be installed at the NR beam port. The present work is the analyses of the proposed irradiation room in view of irradiation safety. To do this, using MCNPX Monte Carlo code the proposed room structure and neutron beam catcher have been simulated and then neutron/gamma dose have been calculated at different locations around the room walls. The results show that the designed room and beam catcher have good performance in case of reduce the total neutron and gamma dose to 10 μSv/h which is desired value in view of radiation protection issue.

In this study, distribution of dose rate around the nuclear gauge device MC-1DR which located in shahrekord university was simulated by MCNPX code and was compared whit the measured values. Due to the asymmetry of device and neutron and gamma source positions, the dose rates were determined at a distance of 5, 30 and 100 cm in different directions. Base on the complex geometry of the inside of device, there are discripency between measured and simulated results in the some points. In general, the values show when the gamma source is positioned in safe mode. The maximum and minimum of dose rate are in below and back of the device. Also, in the left side neutron dose and in the right side gamma dose is greatest. Finally, for safe operating one hour is at most recommended at a distance of 1m in compare with standard threshhold, 12mrem per day.

The purpose of this study was to investigate the effect of voltage ripple produced by X-ray machines on the quality of X-ray rays. For this purpose, the energy spectrum from the X-ray tube manufacturer should be simulated in the range of diagnostic radiography and verified by experimental results. The statistical analysis of the results showed that there is no significant difference between simulation results and experimental results. Then the effect of the ripple factor on the energy spectrum was simulated and investigated. In order to investigate the effect of ripple factor on the beam quality, the half-value layer HVL was calculated. The results of this study showed that the voltage fluctuations can reduce up to 56% of the HVL value at 100 kV and about 75% at 80kV voltage.

Previously earthquake prediction in Iran and world was down with various methods. Result was shown that  the 17-25 days before D/R > 0.5     earthquakes ,  the  FD-RC  was varied rapidly large than standard deviation  in  Reyhan Shahr (Zarand-Kerman) hot spring .therefore  the FD-RC is a powerful  parameter for  earthquake prediction.in this research the experimental data of radon of  Reyhan Shahr(Zarand-Kerman)   hot spring in 2011/12/03 -2012/03/02  time interval was taken from Dr.Negarestani et al  KGUT(Kerman Graduate University of Advanced Technology)  earthquake prediction center. Result was shown that the 15-17 days before D/R > 0.5 earthquakes, the FD-TR was varied rapidly large than standard deviation in Reyhan Shahr (Zarand-Kerman) hot spring. Possibility of earthquake prediction was shown that the D/R > 0.5 earthquakes will happen in 20 – 26 Bahman 1395 in Reyhan Shahr (Zarand-Kerman) zone. Therefore the FD-TR is a powerful parameter for earthquake prediction.

Radioactive materials used in nuclear medicine are the radioisotopes or drugs that are identified with radioisotopes. A radiopharmaceutical transfusion is one of diagnostic methods and a radiation value approved should be measured. The nuclear medicine images of diagnostic methods can be made by a gamma camera (γ-camera). In this article, we used MCNPX2.4 for internal dosimetry assessment when the 64Cu radiopharmaceutical is in brain, heart or breast for imaging or treatment.

In this study, to calculate the radioactivity of radioisotopes such as 11C, 15O, 18F and 123,124I in a focus plasma device for the production of medical radioisotopes, the particles of neutrons and protons with cubic targets of Nitrogen, Xenon, Carbon and a solid natural Boron is used. Particles energy are at the input of the program from 1MeV to 10MeV. Also, the total yield coefficient is considered. In this paper, with the help of MCNPX 2.7 code, the amount of flux and deposited energy of neutron and proton particles is calculated for the production of medical radioisotopes in the focal plasma system. According to the obtained results in this study, it is shown that due to the different absorption and resonance cross sections for the considered targets with the energy of the neutrons and the protons are not a linear increase in the flux of particles. But after an energy of approximately 10 MeV, this slope is almost constant due to the constant cross section. Also, the results of this study show that with increasing energy of protons in these targets, the deposited energy of protons increases linearly, but its maximum value for the nitrogen target and its lowest value for the Xenon target due to different interactions cross-section.

In this study, cobalt 60 with a mean energy of (1.25 Mev) was used as a photonic source to investigate the effect of decreasing gamma-ray intensity by materials and polymer composite using MCNPX code. First, we investigated the effect of thickness and density on the reduction of gamma ray intensity, and then the comparison between the lead protector, tungsten and polymer composite in several thicknesses was simulated by MCNPX code and recorded.

Carbon nanostructures via adding to a polymer matrix, while improving the electrical, mechanical and optical properties of the nanocomposites, are widely used in the industry, medicine, and agriculture. The authors presented several investigations on a new type of gamma dosimeter based on the polymer-carbon nanostructures nanocomposite. In this research, the electrical percolation threshold of the Polystyrene-Graphene Oxide nanocomposite (PS/GO) was simulated using the finite element method in different weight percentages. Then, at the experimental phase, various nanocomposites were fabricated in different weight percentages of 0.05, 0.1, 0.5, 1, 2, 3, 5 and 8 via the mixed-solution process. The electrical conductivities of the samples were measured at the room temperature. Finally, the electrical conductivities of the nanocomposites were compared using the FEM simulation and the experimental results, which were estimated as 2.5wt%. The results of this study showed that the finite element method in accordance with the experimental results is a powerful tool to determine the electrical properties of the polymer nanocomposites considering the dosimetric approaches.

Total body irradiation (TBI) is a kind of external beam radiotherapy which is used in conjunction with chemotherapy with the purpose of immunosuppression before bone marrow transplantation. As recommended by AAPM dose distribution uniformity in TBI is very important and dose variation must be within ±10% of prescription dose. Patients treatment geometry for TBI techniques fall into two common categories: parallel-opposed anterior/posterior lateral total body irradiation and parallel-opposed lateral total body irradiation. So far several techniques such as use of tissue compensator, dynamic TBI delivery, etc. where implemented in different department around the world to improve dose distribution uniformity. The propose of this paper is to review several techniques that are used around the world for dose uniformity improvement in parallel opposed anterior posterior TBI.

Todays gamma knife radiosurgery is used widely for treatment of very small brain tumors. In order to investigate accuracy of dosimetry and treatment planning calculations, using Monte Carlo simulation with dedicated code named as beamnrc including non-CT data and CT data options is necessary. The aim of this study is choosing the best options in order to have an accurate tools based on their advantages and disadvantages. In this study, gamma knife unit 4C along with standard water equivalent phantom and EBT3 films were used to obtain dose distributions. Monte Carlo simulation was done with non-CT data and CT data options of the code and their resulting dose were compared.Comparison the calculated and measured dose distributions at X, Y and Z axis showed gamma value below 1 which verified Monte Carlo simulations. Also comparing the dose distributions from both non-CT data and CT data with each other implies that there is no significant difference between two methods. Based on the obtained results using non-CT data and CT data results in the same dose distribution. So for simplicity, using non CT data for regular phantom shapes is preferred.

In this paper we consider a gaseous detector and supposed, because of pass of an ionizing radiation, an electron created inside it. By numerical simulation with monte carlo method and concluding the impacts, scatterings and creation of secondary electrons, we find the trajectory of initial and secondary electrons. Dependence of number of secondary electrons to applied electrical field is investigated. Finally we study the dependence of number of secondary electrons to number of initial electrons created by ionizing radiation (as a measure of the radiation energy), then we try to investigate the possibility of spectrometry by this detectors.

Dose assessment of radiation victims is a key element in medical management of radiation accidents. At high radiation doses (> 6 Gy), dicentric assay lose its efficiency and premature chromosome condensation (PCC) assay was proposed to overcome the restrictions of dicentric assay. It was shown that PCC ring is more suitable and simpler than dicentric for biodosimetry in high radiation doses. In this study, we established standard dose response curve using chemical PCC ring assay on human peripheral blood lymphocytes for X-ray in the dose-interval of 0–10 Gy. The chemical PCC assay performed on Peripheral blood lymphocyte of four healthy donors after irradiation with different doses of X-rays (0-10 Gy) and the IAEA recommendations for production of an in vitro dose response curve were followed. The dose response curve was fitted based on the linear quadratic model and practicable for high radiation doses. Chemical PCC ring assay is applicable with no need to any special equipment even after supra lethal radiation doses.

Melatonin as a compound made by photosynthetic organizers protects them against many environmental stresses. Modulating mechanisms of melatonin, as a protector are not well known in early eukaryotes, such as algae. In this research, the effect of melatonin pre-treatment on evaluating Chlorella vulgaris resistance to gamma ionizing radiation was investigated. The results showed that 100 µM concentrations of melatonin had a significant increase in the growth of algal cell and activity of the ascorbate peroxidase enzyme as one of the components of the enzymatic antioxidant system.

Developing well logging methods will increase the applications of logs related to all the other geology sciences. Well logging curves introduce the essential information to evaluate reservoir characterizations, rock type and also formation fluid properties quantitatively. One of the most important parts of drilling and completion operations which affect making decision about the future planes is based on how logging data is acquired and interpreted precisely and completely. Litho-density logging tool is a nuclear tool measures formation density and indicating its lithology. Therefore the accuracy of tool measurements is significant. Cs137 is the radioactive source inserted in the tool for downhole measurements. Formation density and lithology are measured by gamma ray Compton scattering and photoelectric absorption phenomena respectively. In this approach, simulation of density tool using MCNP code is discussed to figure out the effect of radioactive sourcechr('39')s activity on the output of energy spectrum belongs to shale, sandstone and limestone formations with known densities. The energy spectrum results affect the accuracy of measurements due to either increase or decrease in source activity.

In this study, the energy spectrum of beta-nickel 63 is considered as the total energy spectrum, average energy of the spectrum, and maximum spectral energy for analyzing the penetration depth of silicon. Monte Carlo calculations were carried out using a MCNPX code in a given geometry; in the following, Stopping-Power for electrons with different energies in silicon was calculated using the ESTAR computational code .The results are in good agreement. Also can be used in the design of semiconductor detectors and nuclear dosimeters.

Although using ionizing radiation is an accepted part of medical profession for patients’ treatment and diagnosis, and despite its benefits, from safety point of view, it can pose potential hazards for radiographers. More understanding of exposure can be an effective move towards safety enhancement of staff and reduction of exposure rate. Present study, aimed at the above mentioned goals, was conducted among different radiographer groups working in hospitals. 35 people from different radiographer groups have been studied, including the practitioners in nuclear medicine, radiotherapy, radiology, and CT scan units in 4 Isfahan state hospitals. Checklists and questionnaires were employed for data gathering; and finally the data were analyzed by SPSS software. Obtained results in this study showed that the average of exposure dose within the past 6 months and the past 30 periods among different working groups enjoyed significant difference, in such a way that average of exposure dose within the past 6 months and the past 30 periods in nuclear medicine group was more than the others. 80% of staff in all the groups use film badge during all shift works. 77.1% of studied person have passed training course regarding the safety in working with radiation and 60% of which used protective tools during the work

Charged particles such as protons and carbon ions are an increasing tool in radiation therapy. However, unresolved physical problems prevent optimal performance, including estimating the deposited dose in non-homogeneous tissue, is an essential aspect of optimizing treatment. The Monte Carlo (MC) method can be used to estimate the amount of radiation, but, this powerful computing operation is very expensive, and has the ability to restrict it. In this work, we use basic physics in the form of the Bethe equation to provide a new analytical solution for range, energy and LET of particles. This solution is presented in terms of the functional integral by converting the relativistic harmonics, which allows it to be used at the level of radiotherapy energy (protons 50-350 MeV, carbon ions of 100-600 Mev / a.m.u). The agreement along the path of the particles, with some differences in reaching the path is high. The model presented in an optimization framework for radiation particle radiation is estimated as a rapid method for dose and LET, which is able to account for heterogeneity in electron density and ionization potential.

The use of a device that simulates the actual condition, as well as being equivalent to the human body in materials and tissues, and having the capability of acting like internal motions of human body is necessary to investigate and simulate the motion of a tumor in radiotherapy and to evaluate dosimetrical coverage of the dynamic target. This device which is called the dynamic phantom should simulate and control the tumor and sorrugate motions.In most phantoms, the bone section is either ignored or is from natural bones instead, which lose their quality over time passing, degraded and cause problems. In this study, a composite material similar to the natural bone has been proposed to replace bone tissue in phantom, which is radiotherapically equal to bone tissue.This composite, in contrast to the natural bone, has a longer life, higher strength and the possibility of embedding a variety of dosimeters. It also prepare dosimetrical evaluation of the organs at risk while the treatment of the dynamic tumor.

The main objective of this paper is to use a computational intelligence algorithm for preparing a mapping map that categorizes different patterns of identification of infected areas and changes in radiation pollution. In this paper, the use of the fuzzy inference system has been proposed to determine the degree of radiation contamination in the regions. The study uses ultra-high resolution spectrometry data to detect uranium. The research area includes well-known uranium deposits, including LambaPur-Peddagattu, Chitrial and Koppunuru. The high-resolution Spectrometry data collected for uranium exploration was used to estimate the average absorption rate in the air due to the distribution of females (potassium per cent and uranium and thorium per million) in these areas. Mamdanichr('39')s Fuzzy Inference System has also been used to determine the amount of radiation contamination in each region. The results showed that the efficiency of this method was 76% accurate for the detection of three levels of radiation contamination (no radiation contamination, low radiation exposure, medium radiation and high radiation pollution) and 89% for the overall identification of contaminated areas from non-polluted areas.

When there is limitation access to one side of the structure in industry, one of the best methods for thickness measurements is the Compton scattering. In this paper, due to the statistical behavior of dispersed beam counting, we discuss the accuracy of the measurement for steel thickness. Using the experimental method and comparing it with MCNP simulation code, the accuracy is obtained less than 0.2 mm.

Thyroid scanning is performed to examine the thyroid gland disorders using 99mTc. Through the obtained images can evaluate the size of the thyroid gland, the amount of its removal compared with the salivary gland, examine the thyroid nodules and its performance impairments. According to the scan view, it is possible to differentiate between hot and cold thyroid nodules that need different therapies. The study of cervical mass with expansion to the chest is another major function of thyroid scanning with 131I. In this method, 24 hours after administration of little 131I, imaging of the neck and chest area is performed. The purpose of the present study was to simulate the emitted particles of 131I radiation and 99mTc, calculate and compare the absorbed dose and the S-value of these two radiopharmaceuticals in the thyroid and other organs around the thyroid, such as the lung, skin, and brain. Simulation and computations were performed using the Monte Carlo method by MCNPX software. ORNL MIRD phantom geometry was used for thyroid and other organs geometry. The absorbed dose in the thyroid and other organs increased with increasing activity, but the absorbed dose in the other organs was lower than the thyroid. The results showed that absorbed dose in 99mTc was lower than 131I. Although  131I has a diagnostic function in thyroid diseases, it has its own special disadvantages. 131 I in addition to 364kev high energy of gamma photons radiations, emits high-energy particles of 647 kev beta, too. Secondly, it has a lifetime of 8/05. Therefore, in studies conducted with 131I, the amount of radiation exposure of the thyroid gland and the total body  is very high (about 15,000 rades are the absorbed thyroid dose). On the other hand, the Tc pertechnetate with a lifetime of 6 hours and gamma energy of 140 kilos of electron volts, like 131I, is absorbed by the thyroid gland. Therefore, regarding the advantages of Tc pertechnetate to the 131I radiopharmaceutical , it is possible to use the absorbing of Tc  pertechnetate  as a measure to evaluate thyroid function.

The inverse-square law is used to calculate the radiation flux at difference distances from the source. This law is applied for point source and point detector in vacuum. This research aims to study the point behavior of volumetric source and detector by undertaking required corrections on inverse-square law through the elimination of scattered neutrons contribution. The measurements have been performed with a 20Ci 241Am-Be neutron source, a long counter and appropriate shadow cone. The results confirm that the exponent in inverse-square law should be changed to 1.907±0.010 when volumetric neutron source and detector are used. Also, the neutron source activity has been measured as 17.95 Ci, which is 5.82% less than its nominal value.

HEPA filters serve as important safety device to protect personnel as well as the public and environment from the radiation effects exist in the air, so they are vital devices in the event accidents. DOE provides directions through the use of two technical standards, 3020 and 3025, including QA requirements for the procurement, packaging, shipping and storage of HEPA filters and also provides direction to the QA requirements for HEPA filter inspection and testing, which their media is Fiber-glass and designed according to the FC and FK sections of ASME AG-1 code.  The DOE Technical Standards specifically state that inspection and testing are to be performed at an independent FTF prior to filter installation at DOE nuclear facilities and this is in addition to the inspection and testing performed by the filter manufacturers.By applying a four step QA inspection and testing program in DOE FTF, the rate of annually deficiencies in visual inspections is 3-5 percent and for flow inspections is about 2 percent and these values sustained during 40 years of applying this program which shows this program is efficient. Also a good relation between AG-1 and DOE is performed, caused the requirements of AG-1 be revised according to this program. In this paper firstly HEPA filter structure and its different types is described, then DOE program of  QA inspections and testing of filters is discussed, and finally deficiencies of ventilation system of TRR in comparison to this program is mentioned in results.

with the advancement of nanotechnology, high-atomic materials such as gold nanoparticles can be used to increase the amount of absorbed doses and use this property to eliminate cancer cells. In this study, the dose enhancement factor (DEF) derived from different concentrations of gold nanoparticles was calculated to show an increase in absorbed dose from gold nanoparticles. At first, a tissue equivalent gel (MAGIC-f) was made at the University of Mazandaranchr('39')s Nuclear Physics Laboratory, and the gold nanoparticles were added to the gel with appropriate concentrations. Vials containing gel and nano-gel were exposed to X-ray radiation of 6 MV of Siemens linear accelerator. After reading the gel with the optical CT scan imaging system, the DEF for nanoparticles was estimated for concentrations, 0.1, 3 and 6 mM. Iit was concluded that increasing the concentration of gold nanoparticles would increase the absorbed dose.