Monte Carlo N-Particle eXtended Research Articles

Modeling of N and P-Type of Coaxial Ge Detectors Using MCNPX and the Effect of Dead Layer Variation on Its Response Function

Abstract This study assessed the response function of a p-type and n-type coaxial high-purity germanium (HPGe) detector via Monte Carlo N-Particle eXtended (MCNPX). MCNPX was employed to model the Coaxial Ge detectors, and for a precise simulation, the dimensions of the dead layer of germanium crystals were added. The dead layer was separated into front and lateral surfaces, and the thickness of each dead layer was modeled. In this work, the simulated detectors have been performed at different energy lines using a radioactive source Eu-152 to study the response function of each with dead layer variations for the front dead layer and study the range of relative deviation of the Monte Carlo simulation outputs from the manufactured declared data. The results proved that the n-type coaxial HPGe detector is more sensitive to the dead layer change than the p-type with a thick change of 0.01 mm. This research has significant effects on the efficiencies of the radiation detection systems in the energy range ∼ (120–1410) keV.

Monte Carlo Methods to Simulate the Propagation of the Created Atomic/ Nuclear Particles from Underground Piezoelectric Rocks through the Fractures Before the Earthquakes

Until now, many studies have been performed on particle radiations before or during earthquakes (EQs). Neutron, gamma, electron, proton, and ultra-low frequency (ULF) photons are among the particles, detected during EQs. In our previous study, with the help of piezoelectricity relationships and the elastic energy formula, the Monte Carlo N‐Particle eXtended (MCNPX) simulation code was applied to find the amount of created atomic/nuclear particles, the dominant interactions; and the energy of the particles for various sizes of quartz and granite blocks. In this study, using the MCNPX simulation code, we have estimated the flux of the particles (created from under-stressed granitic rocks) at different distances from the EQ hypocenter inside the fractures, filled with air, water, and CO2. It was found that inside a water-filled fracture, the particles do not show the flux far from the EQ hypocenter. However, inside the gases like air and CO2 with the normal condition density, different types of particles can have a flux far from the source (more than a kilometer) and they might reach themselves to the surface in the case that the EQ hypocenter is very shallow (0­-5 km). However, for deep EQs, it seems that the most detected nuclear particles on the surface should pass via the vacuum-filled fractures and reach the surface. Moreover, it was concluded that the higher the density of the fracture’s filling fluid, the less distance that the particles can have a flux.

Determination of Exposure during Handling of <sup>125</sup>I Seed Using Thermoluminescent Dosimeter and Monte Carlo Method Based on Computational Phantom

Background: The thermoluminescent dosimeter (TLD) and Monte Carlo (MC) dosimetry are carried out to determine the occupational dose for personnel in the handling of <sup>125</sup>I seed sources.Materials and Methods: TLDs were placed in different layers of the Alderson-Rando phantom in the thyroid, lung and also eyes and skin surface. An <sup>125</sup>I seed source was prepared and its activity was measured using a dose calibrator and was placed at two distances of 20 and 50 cm from the Alderson-Rando phantom. In addition, the Monte Carlo N-Particle Extended (MCNPX 2.6.0) code and a computational phantom with a lattice-based geometry were used for organ dose calculations.Results and Discussion: The comparison of TLD and MC results in the thyroid and lung is consistent. Although the relative difference of MC dosimetry to TLD for the eyes was between 4% and 13% and for the skin between 19% and 23%, because of the existence of a higher uncertainty regarding TLD positioning in the eye and skin, these inaccuracies can also be acceptable. The isodose distribution was calculated in the cross-section of the head phantom when the <sup>125</sup>I seed was at two distances of 20 and 50 cm and it showed that the greatest dose reduction was observed for the eyes, skin, thyroid, and lungs, respectively. The results of MC dosimetry indicated that for near the head positions (distance of 20 cm) the absorbed dose rates for the eye lens, eye and skin were 78.1±2.3, 59.0±1.8, and 10.7±0.7 µGy/mCi/hr, respectively. Furthermore, we found that a 30 cm displacement for the <sup>125</sup>I seed reduced the eye and skin doses by at least 3- and 2-fold, respectively.Conclusion: Using a computational phantom to monitor the dose to the sensitive organs (eye and skin) for personnel involved in the handling of <sup>125</sup>I seed sources can be an accurate and inexpensive method.

Comparison of x-ray spectra and mean glandular dose estimations in computed radiography systems for mammography: Simulations versus measurements

To determine the dose given to patients in mammography, it is necessary to evaluate the mean glandular dose (MGD) in the computed radiography (CR) system with a solid-state detector, utilizing polymethyl methacrylate (PMMA) phantoms and spacers to model breast thickness. In the present study, irradiations were performed in a mammograph with a solid-state detector, and simulations were carried out using the Monte Carlo N-Particle Extended (MCNPX) code to give the MGD and spectra for the same conditions. In the mammograph, the glandular dose values were calculated for three anode/filter combinations: molybdenum/molybdenum (Mo/Mo), molybdenum/rhodium (Mo/Rh), and rhodium/rhodium (Rh/Rh). Through acquisitions from kerma values, the spectra were plotted. The MGD results obtained from the mammograph for Mo/Mo, Mo/Rh, and Rh/Rh were below the reported acceptable levels, while the spectra were mostly in accordance with the literature.

Designing Flexible Shield From Lead With High Protection Against X-rays in Operating Room Using Monte Carlo Simulation

Background: The use of non-flexible lead shields in X-ray detection systems requires a high cost, and it is heavy and subject to frequent cracks and fractures. Therefore, the shield should be designed in a way that has less weight, appropriate efficiency in reducing ionizing radiation, and a more reasonable price. Methods: In the operating room using the MCNPX (Monte Carlo N-Particle extended) code, the shield compounds were modeled to determine the optimal composition (with less weight, higher efficiency in removing ionizing radiation, and in more appropriate price), and then the amount of dose was calculated around the lead chamber with 10 cm from the walls of a detection system. Results: Using a weight combination of 80% lead and 20% polyvinyl chloride was a suitable option in the diagnostic radiology domain which has less weight, lower cost, and good efficiency in reducing X-rays compared to lead shields. Conclusion: The amount of dosage rate on all sides outside the chamber became less than 1 µSv/h, indicating that the combination of lead and polyvinyl chloride is suitable for the body of the chamber.

Evaluation of the Effect of Different Nanoparticles on the Mass Attenuation Coefficient of a Shield in Diagnostic Radiology: A Monte Carlo Study

Purpose: This research aimed to estimate the Mass Attenuation Coefficient (MAC) for the various nanoparticles in diagnostic imaging in order to assess and compare the changes in a bulk state.
 Materials and Methods: To Using Monte Carlo N-Particle eXtended (MCNPX) code, nanoparticles were simulated in the target in order to compute the MAC considering the target. The Materials, including Bi, Pb NPs, Pb, W NPs, W, PbO NPs, Bi NPs, Bi2O3 NPs, and WO3 NPs were used in the present study. The gathered data were compared with the theoretical results of the XCOM software for validation.
 Results: The findings demonstrated that the radioprotective characteristics of nanoparticles in comparison to the bulk materials were better. Among all these nanoparticles, the rate of attenuation of tungsten nanoparticles was higher than that of other nanoparticles. On the other hand, the density and attenuation rate of nanoparticles of PbO, Bi2O3 and WO3 were lower than those of nanoparticles Pb, W, and Bi. Therefore, all of the abovementioned nanoparticles were lightweight and their design was more flexible than that of bulk materials.
 Conclusion: It was concluded that the use of nanoparticles in the protective materials considerably increased the radioprotective characteristics in the diagnostic radiography energy range.

Comparison of MCNPX and EGSnrc Monte Carlo Codes in the Calculation of Nano-Scaled Absorbed Doses and Secondary Electron Spectra around Clinically Relevant Nanoparticles

Purpose: Absorbed dose enhancement due to the presence of high atomic number Nanoparticles (NP)s has been estimated and modeled by Monte Carlo (MC) simulation methods. In the current study, two MC codes of Monte Carlo N‐Particle eXtended (MCNPX) and EGSnrc codes were compared by calculation of secondary electron energy spectra and nano-scaled dose values around four types of spherical NPs.
 Materials and Methods: The MC model was composed of a spherical nanoparticle with a diameter of 50 nm and mono-energetic sources of photons with energies of 30,60, and 100 keV. The secondary electrons emitted from the nanoparticle were scored on the nanoparticle surface and the delivered dose to water around the nanoparticle was tallied using concentric shells with a thickness of 25 nm. Four different elements were used as materials of NPs, including Gold, Bismuth, Gadolinium, and Hafnium.
 Results: Our results showed a considerable difference in the number of emitted electrons per incident photon between the two codes. There were also discrepancies between the two codes in the energy spectra of secondary electrons. Calculated radial dose values around NPs in nano-scale had a similar pattern for both codes. However, significant differences existed for some elements.
 Conclusion: It can be concluded that the results of nano-scaled MC modeling for nanoparticle-based radiation therapy are dependent on the code type and its algorithm for electron transport as well as exploited cross-section libraries.

An alternative method for calculation of half-value layers without the knowledge of attenuation coefficient

Radiation protection is crucial for the safe utilization of ionizing radiation and minimizing the harmful effect upon exposure, hence some standards have been defined by some relevant organizations for the safe uses of radiation. One of the parameters relevant to the calculation of gamma ray shielding is the half-value layer (HVL), which is normally calculated using the knowledge of linear attenuation coefficient (μ). In this research, an attempt has been made to directly calculate HVL without the knowledge of μ via Monte Carlo simulation technique. For this purpose, in the Monte Carlo N‐Particle eXtended (MCNPX) code, F1, F5 and Mesh Popul sequences tallies were defined and the optimal structure for the least measurement error was introduced. The MCNPX calculated values showed reasonable agreement with the experimental findings. According to the obtained results, it is suggested that in order to reduce the error of HVL calculations, in exchange for the MCNPX code, the values of the R parameter and the radiation angle of the source should be considered according to the calculations introduced in this plan. Because the results show that by considering the measurement error between 6 and 20%, the code output can be cited in different energy ranges.

Calculation of NaI(Tl) detector efficiency using 226Ra, 232Th, and 40K radioisotopes: Three-phase Monte Carlo simulation study

Abstract Thallium-activated sodium iodide (NaI(Tl)) detectors can be used in gamma cameras, environmental radiation assessments, including radiation emission levels from nuclear reactors, and radiation analysis equipment. This three-phase investigation aimed to model a standard NaI(Tl) detector using the Monte Carlo N-Particle eXtended (MCNPX) general-purpose Monte Carlo simulation techniques. Accordingly, a standard NaI(Tl) detector was designed along with the required properties. Next a validation study of the modelled NaI(Tl) detector has been performed based on the experimental results for absolute detector efficiency values obtained from 226Ra, 232Th, and 40K radioisotopes. Our findings indicate that the obtained absolute detector efficiency values are quite close to used experimental values. Finally, we used the modelled detector for determination of mass attenuation coefficients of Ordinary concrete, Lead, Hematite-serpentine concrete, and Steel-scrap concrete at 186.1, 295.22, 351.93, 609.31, 1120.29, 1764.49, 238.63, 911.2, 2614, and 1460.83 keV gamma-ray energies. Additionally, according to our findings, mass attenuation coefficients obtained from the newly designed detector are compatible with the standard NIST (XCOM) data. To conclude, continuous optimisation procedures are strongly suggested for sophisticated Monte Carlo simulations in order to maintain a high degree of simulation reliability. As a result, it can be concluded that the validation of the simulation model is necessary using measured data. Finally, it can also be concluded that the validated detector models are effective instruments for obtaining basic gamma-ray shielding parameters such as mass attenuation coefficients.

Monte Carlo simulation of the mass attenuation coefficient and effective atomic number of the Eremurus-Rhizophora ssp. particleboard phantom at the mammography energy range

The earlier investigations proved the potential of the Rhizophora spp. as a phantom material. In this study, the mass attenuation coefficients and empirical and theoretical effective atomic numbers of Eremurus-Rhizophora spp. particleboards in three particle sizes and two adhesive treatment levels were evaluated in 16.63–60 keV photon energy range, using theoretical (XCOM: NIST Photon Cross Sections Database Code), simulation (MCNPX: Monte Carlo N-Particle eXtended Code), and experimental methods. The findings of this study were discussed in terms of the various calculating approaches, and also comparing them with raw wood, binderless particleboard, water, and different breast tissues. The mass attenuation coefficients of Eremurus-Rhizophora spp. samples were seriously agreed with young age breast tissue in three determination methods. Although the empirical effective atomic number (Zeff) is independent of photon energy and the theoretical Zeff is affected by it, the empirical and theoretical effective atomic numbers of the Eremurus-Rhizophora spp. particleboards are in good agreement in the mammography photon energy range. Moreover, the results of this study showed that the radiation features of Rhizophora spp. particleboard can be significantly improved by adding the Eremurus spp. root. The results prove the validity of the MCNP method as an alternative to the XCOM and experimental methods for evaluating the attenuation features and effective atomic numbers of the bio-composites.

Estimating Radiotherapy-Induced Secondary Cancer Risk Arising from Brain Irradiation at High Energy: A Monte Carlo Study

Background: The present study aims to determine the whole-body out-of-field photon dose equivalents of high-energy conventional radiation therapy treatment. Also, it is tried to estimate the probability of fatal secondary cancer risk for the susceptible organs using a Monte Carlo (MC) code.
 Materials and methods: An Monte Carlo N-Particle eXtended (MCNPX)-based model of 18-MV Medical Linear Accelerator (LINAC) was created to calculate the out-of-field photon dose equivalent at the locations of fascinating organs in the mathematical female Medical Internal Radiation Dosimetry (MIRD) phantom. Then, the secondary malignancies risk was estimated based on out-of-field doses and radiation risk coefficients according to the National Council of Radiation Protection and Measurements (NCRP).
 Results: The average photon equivalent dose in out-of-field organs was about 3.25 mSv/Gy, ranging from 0.23 to 37.2 mSv/Gy, respectively, for the organs far from the Planning Target Volume (PTV) (Eyes) and those close to the treatment field (rectum). The entire secondary cancer risk for the 60 Gy prescribed dose to isocenter was obtained as 2.9987%. Here, the maximum doses among off-field organs were related to stomach (0.0805%), lung (0.0601%), and thyroid (0.0404%).
 Conclusion: Regarding the estimated values for the probability of secondary cancer risk, it is suggested to perform a long-term follow-up of brain cancer patients regarding the prevalence of thyroid, stomach, and lung cancer after completing the treatment course.

A novel approach for the reduction of aflatoxin in pistachio nuts using experimental and MCNP simulation

Humans are greatly affected by aflatoxin exposure via the dietary intake of pistachio nuts which pose a serious health risk. This study analysed the reduction of aflatoxin-B1 (AFB1) in pistachio using the exposure of gamma-rays of radioactive granite rock (RGR). The AFB1 level was measured using high-performance liquid chromatography (HPLC). Meanwhile, the reduction of AFB1 in pistachios was examined in two steps. In the first step, AFB1 reduction was examined via different masses of pistachio paste placed on the RGR bed which measured approximately 45% of AFB1 reduction. Whereas, 97% reduction in AFB1 was observed in pistachios within the small packets placed on the RGR bed. The second step was designed to investigate the effects of irradiation time and mass of RGR bed on the AFB1 reduction using RGR bed. Based on the results, the increase in AFB1 reduction level was recorded by increasing the γ-ray irradiation time and the mass of RGR. Through the introduction of the pistachio dosimetric AFB1 reduction coefficient (η), an empirical formula was developed to predict the amount of AFB1 reduction in the pistachio by RGR bed. Overall, the consistency level between empirical formulas and practical results is about 84–87%. Meanwhile, the gamma radiation was calculated using the Monte Carlo N Particle eXtended (MCNPX) simulation code. In conclusion, the proposed novel technique demonstrated RGR to be a suitable industrial substrate for safe and hygienic storage of pistachios.

A neutronic investigation of tristructural isotropic‐duplex fuel in a high‐temperature reactor prismatic seed‐and‐blanket fuel block configuration with reduced power peaking

This paper presents the neutronic analysis of a tristructural isotropic (TRISO)-duplex fuel loaded into the high-temperature reactor (HTR) prismatic fuel block designs for a seed-and-blanket (S&B) reactor model. The selected fuel block design is the simplified model of a representative fuel configuration for the gas turbine–module helium reactor (GT-MHR) and high-temperature engineering test reactor (HTTR). The duplex fuel pellet comprises of UO2 as the seed and ThO2 as the blanket and uses the TRISO fuels. Monte Carlo N-particle extended (MCNPX) simulations were conducted to analyze the power and neutron flux distributions of the reference fuel block and that of the TRIOS-loaded duplex fuel designs. Standard neutronic parameters were computed, such as burnup-dependent infinite multiplication factors (kinf), average block spectrum, and fissile inventory. It was discovered that while the S&B model could attain the highest burnup and the longest cycle duration, its power peaking was more than threefolds than that of the reference model at the beginning of the cycle (BOC). The results also show that the “Duplex-2” model could reduce its power peaking by 24% through the integration of S&B arrangement with the duplex rods. It is noted that spatial distribution of the UO2 seed and 235U enrichment are the chief determining factors in the power optimization of the heterogeneous HTR prismatic fuel blocks loaded with thorium.

Modeling of x-ray attenuation on SS316 metal foam shielding using MCNPX

The use of lead plates as shielding to X-ray radiation has the disadvantages of being heavier and not environmentally friendly when compared to other materials. Currently, metal foam technology is developed that is expected to be close to the ability of lead plate as a radiation shielding that is lighter and environmentally friendly. This research was conducted by modeling SS316 stainless steel plates and foams for the thickness of 5-40 mm, the diameter of 2, 3, and 4 mm at an energy of 0.1, 0.662, and 1 MeV. The metal plate and foam modeling were carried out using the program of Monte Carlo N-Particle eXtended (MCNPX). Calculation of X-ray attenuation from the modeling results shows that SS316 stainless steel performance both in the form of plates and metal foams is almost equivalent to pure lead. From these results, it can be concluded that metal foam technology is expected to be a lighter and environmentally friendly lead-in radiation retaining material.

Effect of Humeral Locking Plate System on Absorbed Energy in Breast Tissue with Different Radiological Energies Using MCNPX Code

Abstract This study aimed to investigate the radiation absorption effects of the humeral locking plate system on breast tissue in different radiological energy ranges using MCNPX (Monte Carlo N-Particle Extended Transport Code System, version 2.6.0) general-purpose Monte Carlo code. In the present study, a novel MCNPX simulation setup has been generated considering the elemental mass fractions of investigated materials and biological structures. The validation of MCNPX simulation geometry used in this study has been provided by comparing the results with standard XCOM data for mass attenuation coefficients of Ti-6Al-4V alloy. A very good match between XCOM and MCNPX data has been achieved. Therefore, validated simulation input has been employed for further investigations of the present study. Our findings showed that the amount of absorbed energy that was measured in breast tissue with the humeral locking plate was higher than that without humeral locking plate for the same radiological energy value. Our results would be beneficial to use the present simulation technique and mass attenuation coefficients for radiation physics and medical applications, especially for assessments of breast cancer risk.

COMPARISON OF SPECTRA AND MEAN GLANDULAR DOSE WITH TUBE VOLTAGES USED IN DIGITAL BREAST TOMOSYNTHESIS FROM SIMULATED, METROLOGICAL AND CLINICAL CASES.

Digital breast tomosynthesis (DBT) is a screening and diagnostic modality that acquires images of a breast at multiple angles during a short scan. The Selenia Dimensions (Hologic, Bedford, Mass) DBT system can perform both full-field digital mammography and DBT. The system acquires 25 projections over a 15° angular range (from -7.5° to +7.5°). X-ray spectroscopy is generally linked with a high-resolution semiconductor detector through a correction to its energy response function. The energy spectrum describes the radiation field, in which several quality parameters can be extracted, such as the effective energy, half-value layer and exposure. X-ray spectroscopy is usually performed with solid-state semiconductor detectors. Radiation dose is a concern in mammography, as the current protocols recommend that medical physicians evaluate mean glandular dose (MGD) as a part of service quality control. Studies are needed for radiation dose optimization from tomosynthesis patients. The COMET metrological X-ray tube, considered as with a constant potential and cooled, has proved to be a crucial tool in order to obtain the high energy resolution for low-energy radiographs in mammography. The Monte Carlo method, through Monte Carlo N-Particle eXtended (MCNPX), was proven to be an essential tool for image formation and posterior analysis of the deposited dose from breast simulators and radiographic contrast evaluation, for several anode/filter combinations. The purpose of this work was to assess the MGD and spectra in slabs of polymethyl methacrylate (PMMA) and breast equivalent thicknesses using four experiments with a Hologic Selenia Dimensions mammography X-ray tube with multimeter, a spectrometer (only for spectra, in this case), a metrological X-ray tube with a multimeter, and the MCNPX code. References indicate that the real conditions for a mammography X-ray tube that conducts tomosynthesis include tube voltages of 26, 29, 30 and 33kVp. Taking into account several thicknesses of PMMA, both the MGD and spectral results were in accordance with the references. Most of the spectra were in accordance with the references, showing that the resources used in the experiments can evaluate the energy level received by a patient. The MGD values were lower than those in the references from 30 to 50mm PMMA, and the data can be used for improvements in the detectors used in the Laboratory of Metrology in the State of Rio de Janeiro University, Brazil. Additionally, in the future, optimization of image quality can be performed for both semiconductors and mammography X-ray equipment.

Structural and nuclear shielding qualities of B2O3–PbO–Li2O glass system with different Ag2O substitution ratios

Structural and gamma-ray/neutron radiation protection characteristics of (50-x)B2O3 +40PbO+10Li2O + xAg2O (x = 0, 2.5, 5, 7.5, 10 wt fraction (wt) %) glass system encoded BPLA glasses have been extensively surveyed. The structure of the prepared glasses has been investigated over X-Ray Diffraction (XRD) patterns. With the addition of Ag2O, the glasses' amorphous structures have not deteriorated, and the density and molar volumes of BPLA glasses inversely varied. Mass attenuation coefficients (MAC) for BPLA glasses have been generated enjoying Monte Carlo N-Particle- Extended (MCNPX) simulation code for the broad gamma-ray energy range of 0.015–15 MeV. It has been observed that the MAC values of the BPLA10 glass are higher than the others at all energy levels. Furthermore, while Ag2O concentration in the glass has increased, Half Value Layer (HVLs) Energy Absorption and Exposure Buildup Factors (EABFs, and EBFs) have dropped off for photon energies which are important for shielding applications. It has been noticed from these results that owning higher equivalent atomic number (Zeq) values, BPLA10 can be employed effectively for gamma-ray attenuation. Lastly, the results found for fast neutron removal cross-sections (∑R) have shown that the Ag2O additive grew the density and ∑R values of the glasses. As with photon shielding, it is observed that BPLA10 glass can be more preferable in terms of neutron retention according to other glasses. One can conclude that Ag2O enhances the shielding capacity of the glass system studied and BPLA10 glass can be considered as a good material for gamma-rays/fast neutrons shielding.

Conceptual Shield Design for Boron Neutron Capture Therapy Facility Using Monte Carlo N-Particle Extended Simulator with Kartini Research Reactor as Neutron Source

The research aims to measure the radiation dose rate over the radiation shielding which is made of paraffin and aluminium and to determine the best shield material for the safety of radiation workers. The examination used MCNP (Monte Carlo N-Particle) simulator to model the BNCT neutron source and the shield. The shield should reduce radiation to less than the dose limit of 10.42 µSv/h, which is assumed to be the most conservative limit when the duration of workers is 1920 h. The first design resulted in a radiation dose rate which was still greater than the limit. Therefore, optimization was done by adding the lead on the outer part of the shield. After optimization by adding the lead with certain layers, the radiation dose rate decreased, with the largest dose being 57.60 µSv/h. Some locations over the limit could be overcome by other radiation protection aspects such as distance and time. The paraffin blocks were covered by aluminium to keep the shield structure. The lead was used to absorb the gamma ray which resulted from the interaction between the neutrons and aluminium.

Lead-free transparent shields for diagnostic X-rays: Monte Carlo simulation and measurements.

In recent years, the preference for using lead-free radiation protection shields has increased because of concerns regarding lead poisoning and leakage. In medical and research laboratories, glass shields are preferred because of their transparency. In this study, various glass shields were examined and compared based on the international standards. One commercially available lead-based shield, four recently studied shields, and three new lead-free shields were considered, and their shielding factors were calculated. We presented three glasses based on borate, phosphate, and silicate compounds, which were named Ir1, Ir2, and Ir3, respectively. Based on the International Electrotechnical Commission standard (IEC 61331), the air-kerma ratios (attenuation ratios) and lead equivalent values were derived using Monte Carlo N-Particle eXtended (MCNPX) calculations, and mass attenuation coefficients and effective atomic numbers (Zeff) of all the shields were obtained from XCOM database, in the diagnostic X-ray energy range of 40-120keV. In addition, some measurements were performed for the reference (lead-based) glass to validate the simulations. The above-mentioned factors for silicate-bismuth-based (Ir3) and borosilicate-barium-based (Tu) glasses were found to be higher than the others and comparable to those of commercially available lead-based glass. In conclusion, Ir3 and Tu glasses were found to be the preferred lead-free transparent shields in the diagnostic X-ray energy range.

The impact of gadolinium on the reactor production of 153Sm

Radioisotopes represent major sources of ionizing radiation, not least for use in medical applications, brachytherapy and nuclear medicine included. In this, the nuclear reactor is the main source of β- - γ emitting isotopes, an example product being 153Sm used in the treatment of pain arising from bone metastases. Present analysis relates to the potential of gadolinium neutron capture reaction, its impact on reactor production of radioisotopes and the proliferation resistant potential of thorium fuel cycle. A comparative analysis has been made of the impact of gadolinium on the production of 153Sm by UO2 and (Th, U) O2 fuels in a Westinghouse small modular reactor. Five fuel assemblies were investigated: one containing no gadolinium, the other four containing 16, 24, 34 or 44 gadolinium fuel rods. The code Monte Carlo N-Particle eXtended (MCNPX) integrated with the CINDER90 burn-up code was used for calculations. In the production of 153Sm the same trend is followed for the fuels containing gadolinium, increasing significantly with the number of gadolinium fuel rods. Zero production results from fuel assemblies without gadolinium. The concentration of 153Sm increases significantly with burn-up, indicating that gadolinium has a positive impact on the production of 153Sm.