General Purpose Monte Carlo Code Research Articles

Changes in molecular conformation and electronic structure of DNA under 12C ions based on first-principles calculations

Fundamental studies on the interaction between ionizing radiation and deoxyribonucleic acid (DNA) can contribute to understanding the biological effects of radiation protection and radiotherapy. The transitional process from ionized and excited DNA to DNA lesions, such as strand breaks and base lesions, remains uncertain. In this study, we focused in particular on the direct effects of radiation on DNA and calculated the number of ionizations induced within 10 base pairs of DNA using the Particle and Heavy Ion Transport code System, a general-purpose Monte Carlo code for radiation transport. The electronic structure of the ionized DNA is subsequently calculated using the first-principles calculation software OpenMX. Through these calculations, we successfully identified the redox site of DNA and quantitatively evaluated the amount of molecular bond cleavage. The scientific findings of this study will elucidate the cleavage mechanism of DNA in the high-density radiation field realized by ion beam irradiation by determining the redox sites.

Enhancing mechanical, physical, radiation attenuation properties in alumino-barium-titanium-calcium-lithium glasses for nuclear applications: The pivotal role of TiO2additives

This study focused on enhancing key material properties of Alumino-Barium-Titanium-Calcium-Lithium glasses for nuclear applications, specifically by integrating increasing amounts of TiO2 additives. Utilizing the MCNPX general-purpose Monte Carlo code, version 2.7.0, the research aimed to ascertain Transmission Factor (TF) values across a spectrum of well-known radioisotope energies. This analysis was conducted for glass samples with varying thicknesses, ranging from 0.5 cm to 3 cm. The study also delved into the gamma-ray shielding characteristics of these glasses at energy levels between 0.015 and 15 MeV, uncovering notable findings. In exploring the T1 to T12 glass system, various physical and optical methods were employed to measure key parameters like glass density (ρglass), molar volume (Vm), oxygen molar volume (OMV), and oxygen packing density (OPD). A significant outcome of this research was the observation that with an increase in TiO2 content, there was a corresponding rise in glass density, from 3.727 g/cm3 to 3.825 g/cm3. Furthermore, the study noted alterations in the physical and mechanical properties of the glasses. Most notably, the T12 glass sample, which contained the highest concentration of TiO2, exhibited superior gamma-ray shielding properties compared to the other glass compositions analysed.

The effect of two-part dead layer thickness on the efficiency curve of a P-TYPE HPGe detector calculated using the Monte Carlo method

The full energy peak efficiency (FEPE) of a gamma ray obtained from experimental studies using a high purity germanium detector (HPGe) can be easily reproduced by simulating the experiment in any desired geometry and energy range with Monte Carlo (MC) simulation methods. However, the dead layer (DL) thickness, which is one of the geometric information provided by the manufacturer, changes over time and affects the efficiency calculation, so it must be determined and compared with the initial value for the increases during the life-time of the detector. In this study, the general-purpose MC code PHITS and the special-purpose MC code GESPECOR were used to simulate a p-type HPGe detector and to determine the thickness of the frontal and lateral dead layer. As a result, it was observed that the frontal dead layer thickness increased from 0.7 mm to 1.075 mm, and the lateral dead layer thickness increased from 0.7 mm to 1.7 mm after 7 years of operation of the detector. This increase in dead layer thickness causes both a decrease in the detector's active volume and a decrease in efficiency in the problematic low-energy region.

Exploring a monotonically non-decreasing behavioral function on shielding properties and transmission factors in borosilicate glasses through heavy metal oxide reinforcement

In this study, the gamma-ray attenuation properties and transmission factors of different types of glasses formed in the (B2O3)x[(ZnO)0.60-(SiO2)0.40]1-x system as a function of increasing heavy metal oxide ratio were investigated. For transmission factor and absorption coefficient calculations, the general-purpose Monte Carlo code MCNPX (2.7.0) was used. In addition, the Phy-X/PSD code is used to obtain some important gamma-ray shielding parameters. According to the obtained results, increasing the quantity of heavy metal oxide in the glass composition favorably affected the examined parameters and lowered the transmission factor and buildup factor values. Furthermore, it has been observed that parameters such as the absorption coefficient and the effective atomic number increase as a function of increasing heavy metal oxide contribution into glass configuration. The lowest half value layer values are reported for S1 sample. For example, half value layers are reported as 3.313 cm, 3.449 cm, 3.764 cm, 3.625 cm, 3.515 cm for S1, S2, S3, S4, and S5 sample at 1 MeV photon energy, respectively. According to the transmission factor values, S1 is the sample that allows the least transmission of primary photons and demonstrates the highest resistance. It was observed that the minimal exposure build-up factor and energy absorption buildup factor values for the S1 sample at 10 mean free paths. It can be concluded that the heavy metal oxide additive applied to borosilicate glasses creates a non-decreasing functional behavior effect that is monotonic in gamma-ray attenuation parameters.

Introduction to the Monte Carlo dose engine COMPASS for BNCT.

The Monte Carlo method is the most commonly used dose calculation method in the field of boron neutron capture therapy (BNCT). General-purpose Monte Carlo (MC) code (e.g., MCNP) has been used in most treatment planning systems (TPS) to calculate dose distribution, which takes overmuch time in radiotherapy planning. Based on this, we developed COMPASS (COMpact PArticle Simulation System), an MC engine specifically for BNCT dose calculation. Several optimization algorithms are used in COMPASS to make it faster than general-purpose MC code. The parallel computation of COMPASS is performed by the message passing interface (MPI) library and OpenMP commands, which allows the user to increase computational speed by increasing the computer configurations. The physical dose of each voxel is calculated for developing a treatment plan. Comparison results show that the computed dose distribution of COMPASS is in good agreement with MCNP, and the computational efficiency is better than MCNP. These results validate that COMPASS has better performance than MCNP in BNCT dose calculation.

Implementation of Visible monkey into general-purpose Monte Carlo codes: MCNP, PHITS, and Geant4

Recently, a new monkey computational phantom, called Visible Monkey, was developed for non-ionizing radiation studies in animal research. In this study, we extended its applications to ionizing radiation studies by implementing the voxel model of the Visible Monkey into three general-purpose Monte Carlo (MC) codes: MCNP6, PHITS, and Geant4. The implementation work for MCNP and PHITS was conducted using the LATTICE, UNIVERSE, and FILL cards. The G4VNestedParameterisation class was used for Geant4. Then, organ dose coefficients (DCs) for idealized photon beams in the antero-posterior direction were calculated using the three codes and compared, showing excellent agreement (differences <3%). Additionally, organ DCs in other directions (postero-anterior, left-lateral, and right-lateral) were calculated and compared with those of the newborn and 1-year-old reference phantoms. Significant differences were observed (e.g., the stomach DC of the monkey was 5-fold greater than that of the 1-year-old phantom at 0.03 MeV) while the differences tended to decrease with increasing energy (mostly <20% at 10 MeV). The results of this study allows conducting MC simulations using the Visible Monkey to estimate organ-level doses, which should be valuable to support/improve monkey experiments involving ionizing radiation exposures.

Designing a Lead-free and high-density glass for radiation facilities: Synthesis, physical, optical, structural, and experimental gamma-ray transmission properties of newly designed barium-borosilicate glass sample

We report the design, synthesis, optical, structural, and gamma-ray attenuation properties of a newly developed Lead-free and high-density borosilicate glass sample for its potential applications in medical and industrial radiation facilities. A barium-borosilicate glass sample (BSBaZn) was designed and synthesized using nominal composition of 7B2O3-50SiO2-38ZnO-5BaO. The FTIR spectrum of the BSBaZn is revealed four fundamental regions. These regions are 400–620 cm−1, 620–770 cm−1, 800–1210 cm−1, and 1210–1500 cm−1. Transmittance rate in the wavelength range of 350–1100 nm is reported as 80 %. A high-purity Germanium (HPGe) detector along with an energetic 133Ba radioisotope is also utilized for experimental gamma-ray transmission studies. Various fundamental gamma-ray shielding parameters of BSBaZn are determined and accordingly compared with many other glass shields. MCNPX (version 2.7.0) general purpose Monte Carlo code is utilized for gamma-ray transmission factor (TF) values. The results showed that the synthesized BSBaZn sample has promising structural, optical, and physical properties in addition to promising gamma-ray attenuation properties. The high transparency of BSBaZn along with its high-density may be considered as an important selection criterion for its implementation in protection purposes in medical and industrial radiation facilities, where the source and patients monitoring play a significant role.

Toward the strengthening of radioprotection during mammography examinations through transparent glass screens: A benchmarking between experimental and Monte Carlo simulation studies.

A lead-acrylic protective screen is suggested to reduce radiation exposure to the unexposed breast during mammography. The presence of toxic lead in its structure may harm the tissues with which it comes in contact. This study aimed to design a CdO-rich quaternary tellurite glass screen (C40) and evaluate its efficiency compared to the Lead-Acrylic protective screen. A three-layer advanced heterogeneous breast phantom designed in MCNPX (version 2.7.0) general-purpose Monte Carlo code. Lead acrylic and C40 shielding screens were modeled in the MCNPX and installed between the right and left breast. The reliability of the absorption differences between the lead acrylic and C40 glass were assessed. The results showed that C40 protective glass screen has much superior protection properties compared to the lead acrylic protective screen. The amount of total dose absorbed in the unexposed breast for C40 was found to be much less than that for lead-based acrylic. The protection provided by the C40 glass screen is 35-38% superior to that of the Lead-Acrylic screen. The C40 offer the opportunity to avoid the toxic Pb in the structure of Lead-Acrylic material and may be utilized for mammography to offer superior radioprotection to Lead-Acrylic and significantly lower the dose amount in the unexposed breast. It can be concluded that transparent glass screens may be utilized for radiation protection purposes in critical diagnostic radiology applications through mammography.

Comparative analysis on application conditions of indium (III) oxide-reinforced glasses in nuclear waste management and source transportation: A Monte Carlo simulation study

This study's primary objective is to provide the preliminary findings of novel research on the design of Indium (III) oxide-reinforced glass container that were thoroughly developed for the purpose of a nuclear material container for transportation and waste management applications. The shielding characteristics of an Indium (III) oxide-reinforced glass container with a certain elemental composition against the 60Co radioisotope was thoroughly evaluated. The energy deposition in the air surrounding the designed portable glass containers is measured using MCNPX general-purpose Monte Carlo code. Simulation studies were carried out using Lenovo-P620 workstation and the number of tracks was defined as 108 in each simulation phase. According to results, the indium oxide-doped C6 (TZI8) container exhibits superior protective properties compared to other conventional container materials such as 0.5Bitumen-0.5 Cement, Pb Glass composite, Steel-Magnetite concrete. In addition to its superiority in terms of nuclear safety, it is proposed that the source's simultaneous observation and monitoring, as well as the C6 (TZI8) glass structure's transparency, be underlined as significant advantages. High-density glasses, which may replace undesirable materials such as concrete and lead, provide several advantages in terms of production ease, non-toxic properties, and resource monitoring. In conclusion, the use of Indium (III) oxide-reinforced glass with its high transparency and outstanding protection properties may be a substantial choice in places where concrete is required to ensure the safety of nuclear materials.

On the mystery of the multi-muon flux at the TeV cosmic-ray energy range

Current Monte Carlo simulations do not provide a good description of the muon component of extensive air showers. Many air shower experiments report discrepancies between their data and Monte Carlo predictions, ranging from the TeV scale up to the highest energies. In these proceedings, we address the seasonal variation of the multi-muon events observed by the NOvA Near Detector (ND). For our studies, we use the general-purpose Monte Carlo code FLUKA to treat the transport and interaction of the air-shower particles in the atmosphere and other media. Our design considers a multilayered atmosphere and a layered underground approximated to match the NOvA ND location and detector geometry. Our atmospheric model uses air densities for winter and summer calculated from the temperature and geopotential information for the pressure levels given by the European Center for Medium-Range Weather Forecasts (ECMWF) datasets in situ. Understanding the multi-muon flux at the cosmic ray high-energy range may lead to a better description of the muon production mechanisms in ultra-high-energy extensive air showers. In addition, it can help to improve future Monte Carlo codes or hint at new physics processes or interactions.

A critical evaluation on nuclear safety properties of novel cadmium oxide-rich glass containers for transportation and waste management: Benchmarking with a reinforced concrete container

We examine the nuclear safety properties of a newly designed cadmium oxide-rich glass container for nuclear material to a bitumen-reinforced concrete container. Individual transmission factors, detector modelling, and energy deposition (MeV/g) in the air are calculated using MCNPX (version 2.7.0) general purpose Monte Carlo code. Two container configurations are designed with the material properties of cadmium dioxide-rich glass and Concrete + Bitument in consideration. First, individual transmission factors for 60Co and 137Cs radioisotopes are calculated. To evaluate potential environmental consequences, energy deposition amounts in the air for 60Co and 137Cs are also determined. The minimum gamma-ray transmission rates for two container types are reported for a cadmium dioxide-rich glass container. In addition, the quantity of energy deposition is varied depending on the container type, with a lower value for cadmium dioxide-rich glass container. The 40% cadmium dioxide-doped glass container provides more effective safety than the Cement + Bitumen container, according to the overall findings. In conclusion, the utilization of cadmium dioxide-doped glass material along with its high transparency and advanced material properties may be a significant and effective option in areas where concrete is required to assure the safety of nuclear materials.

Characterisation of two p-type HPGe detectors by measurements and Monte Carlo simulations

This study aims to characterize two p-type HPGe detectors with the same geometric properties, which were put into operation 9 years ago by determining the optimum dead layer thickness. Experimental efficiencies were measured in the energy range of 81–1408 keV using point sources. As a result of simulations using general-purpose and specific-purpose Monte Carlo codes, optimized dead layer thicknesses, which were stated to be <1 mm after production, were 1.19 mm for Det1 and 1.71 mm for Det2. Experimental and Monte Carlo full energy peak efficiency values calculated with optimized dead layer thicknesses have the greatest difference of 5 %. The two detectors were characterized as a result of the optimized thicknesses determined by the Monte Carlo simulations. It was concluded that one reason for the variation in the dead layer specified by the manufacturer is vacuum degradation.

Impact of the Lorentz force on electron track structure and early DNA damage yields in magnetic resonance-guided radiotherapy

Magnetic resonance-guided radiotherapy (MRgRT) has been developed and installed in recent decades for external radiotherapy in several clinical facilities. Lorentz forces modulate dose distribution by charged particles in MRgRT; however, the impact of Lorentz forces on low-energy electron track structure and early DNA damage induction remain unclear. In this study, we estimated features of electron track structure and biological effects in a static magnetic field (SMF) using a general-purpose Monte Carlo code, particle and heavy ion transport code system (PHITS) that enables us to simulate low-energy electrons down to 1 meV by track-structure mode. The macroscopic dose distributions by electrons above approximately 300 keV initial energy in liquid water are changed by both perpendicular and parallel SMFs against the incident direction, indicating that the Lorentz force plays an important role in calculating dose within tumours. Meanwhile, DNA damage estimation based on the spatial patterns of atomic interactions indicates that the initial yield of DNA double-strand breaks (DSBs) is independent of the SMF intensity. The DSB induction is predominantly attributed to the secondary electrons below a few tens of eV, of which energy deposition patterns are not considerably affected by the Lorentz force. Our simulation study suggests that treatment planning for MRgRT can be made with consideration of only changed dose distribution.

Experimental validation of Monte Carlo dosimetry for therapeutic beta emitters with radiochromic film in a 3D-printed phantom.

Validation of dosimetry software, such as Monte Carlo (MC) radiation transport codes used for patient-specific absorbed dose estimation, is critical prior to their use in clinical decision making. However, direct experimental validation in the clinic is generally not performed for low/medium-energy beta emitters used in radiopharmaceutical therapy (RPT) due to the challenges of measuring energy deposited by short-range particles. Our objective was to design a practical phantom geometry for radiochromic film (RF)-based absorbed dose measurements of beta-emitting radionuclides and perform experiments to directly validate our in-house developed Dose Planning Method (DPM) MC code dedicated to internal dosimetry. The experimental setup was designed for measuring absorbed dose from beta emitters that have a range sufficiently penetrating to ∼200 μm in water as well as to capture any photon contributions to absorbed dose. Assayed 177 Lu and 90 Y liquid sources, 13-450 MBq estimated to deliver 0.5-10 Gy to the sensitive layer of the RF, were injected into the cavity of two 3D-printed half-cylinders that had been sealed with 12.7 μm or 25.4 μm thick Kapton Tape. A 3.8 × 6 cm strip of GafChromic EBT3 RF was sandwiched between the two taped half-cylinders. After 2-48 h exposures, films were retrieved and wipe tested for contamination. Absorbed dose to the RF was measured using a commercial triple-channel dosimetry optimization method and a calibration generated via 6 MV photon beam. Profiles were analyzed across the central 1 cm2 area of the RF for validation. Eleven experiments were completed with 177 Lu and nine with 90 Y both in saline and a bone equivalent solution. Depth dose curves were generated for 177 Lu and 90 Y stacking multiple RF strips between a single filled half-cylinder and an acrylic backing. All experiments were modeled in DPM to generate voxelized MC absorbed dose estimates. We extended our study to benchmark general purpose MC codes MCNP6 and EGSnrc against the experimental results as well. A total of 20 experiments showed that both the 3D-printed phantoms and the final absorbed dose values were reproducible. The agreement between the absorbed dose estimates from the RF measurements and DPM was on average -4.0% (range -10.9% to 3.2%) for all single film 177 Lu experiments and was on average -1.0% (range -2.7% to 0.7%) for all single film 90 Y experiments. Absorbed depth dose estimates by DPM agreed with RF on average 1.2% (range -8.0% to 15.2%) across all depths for 177 Lu and on average 4.0% (range -5.0% to 9.3%) across all depths for 90 Y. DPM absorbed dose estimates agreed with estimates from EGSnrc and MCNP across the board, within 4.7% and within 3.4% for 177 Lu and 90 Y respectively, for all geometries and across all depths. MC showed that absorbed dose to RF from betas was greater than 92% of the total (betas + other radiations) for 177 Lu, indicating measurement of dominant beta contribution with our design. The reproducible results with a RF insert in a simple phantom designed for liquid sources demonstrate that this is a reliable setup for experimentally validating dosimetry algorithms used in therapies with beta-emitting unsealed sources. Absorbed doses estimated with the DPM MC code showed close agreement with RF measurement and with results from two general purpose MC codes, thereby validating the use of this algorithms for clinical RPT dosimetry.

Heavy metal oxide added glassy portable containers for nuclear waste management applications: In comparison with reinforced concrete containers

This study aimed to investigate the protective properties of Bi2O3 heavy metal oxide-doped glassy portable containers and the effect of reinforcement amount on these properties using the MCNPX (version 2.6.0) general-purpose Monte Carlo code. Accordingly, 60Co and 137Cs radioisotopes were defined as point isotropic radioactive sources to be transported with the newly designed containers. Four containers with different heavy metal oxide additives varying between 5% and 20% were designed and the deposited energy (MeV/g) values in the air were calculated for both 60Co and 137Cs radioisotopes. According to the findings of the first phase of the investigation, the sample (S4) with a 20% Bi2O3 additive ratio showed the highest protective properties and the least amount of deposited energy amount in the air. In the second and benchmarking phase of the investigation, we compared the amount of deposited energy in the air for the superior S4 glass container and a concrete container with a high amount of bitumen additive. The findings demonstrated that the S4 portable glass container with a 20% Bi2O3 reinforcement provided significantly lower deposited energy in the air and therefore greater nuclear safety than the concrete container. Heavy metal oxide-doped glass may be considered a viable choice for nuclear waste management and transportation operations due to its nuclear safety properties and superior physical, optical, and mechanical capabilities in comparison with concrete.

Diagnostic and therapeutic radioisotopes in nuclear medicine: Determination of gamma-ray transmission factors and safety competencies of high-dense and transparent glassy shields

Abstract We present the findings of an extensive examination on newly designed CdO-rich and transparent glass shields for nuclear medicine facilities in lieu of traditional and unfavorable materials, such as lead and concrete. Gamma-ray transmission factors of newly designed glass shields are determined using a variety of diagnostic, therapeutic, and research radioisotopes, including 67Ga, 57Co, 111In, 201Tl, 99mTc, 51Cr, 131I, 58Co, 137Cs, 133Ba, and 60Co. A general-purpose Monte Carlo code MCNPX (version 2.7.0) is used to determine the attenuation parameters of different material thicknesses. Next, the findings are compared using a standard concrete shielding material. The results indicate that adding more CdO to the glass composition improves the overall gamma-ray attenuation properties. As a result, among the heavy and transparent glasses developed, the C40 sample containing 40% CdO exhibited the best gamma-ray absorption properties against all radioisotopes. Furthermore, the gamma-ray absorption characteristics of this created high-density glass were shown to be better to those of a standard and heavy concrete sample. It can be concluded that the newly developed CdO-rich and transparent glass sample may be used in medical radiation fields where the radioisotopes examined are used in daily clinical and research applications.

Monte Carlo characterization of high atomic number inorganic scintillators for in vivo dosimetry in 192 Ir brachytherapy.

There is increased interest in in vivo dosimetry for 192 Ir brachytherapy (BT) treatments using high atomic number (Z) inorganic scintillators. Their high light output enables construction of small detectors with negligible stem effect and simple readout electronics. Experimental determination of absorbed-dose energy dependence of detectors relative to water is prevalent, but it can be prone to high detector positioning uncertainties and does not allow for decoupling of absorbed-dose energy dependence from other factors affecting detector response. To investigate which measurement conditions and detector properties could affect their absorbed-dose energy dependence in BT in vivodosimetry. We used a general-purpose Monte Carlo (MC) code PENELOPE for the characterization of high-Z inorganic scintillators with the focus on ZnSe ( ) Z. Two other promising media CsI ( ) and Al2 O3 ( ) were included for comparison in selected scenarios. We determined absorbed-dose energy dependence of crystals relative to water under different scatter conditions (calibration phantom 12 × 12 × 30cm3 , characterization phantoms 20 × 20 × 20 cm3 , 30 × 30 × 30 cm3 , 40 × 40 × 40 cm3 , and patient-like elliptic phantom 40 × 30 × 25 cm3 ). To mimic irradiation conditions during prostate treatments, we evaluated whether the presence of pelvic bones and calcifications affect ZnSe response. ZnSe detector design influence was alsoinvestigated. In contrast to low-Z organic and medium-Z inorganic scintillators, ZnSe and CsI media have substantially greater absorbed-dose energy dependence relative to water. The response was phantom-size dependent and changed by 11% between limited- and full-scatter conditions for ZnSe, but not for Al2 O3 . For a given phantom size, a part of the absorbed-dose energy dependence of ZnSe is caused not due to in-phantom scatter but due to source anisotropy. Thus, the absorbed-dose energy dependence of high-Z scintillators is a function of not only the radial distance but also the polar angle. Pelvic bones did not affect ZnSe response, whereas large and intermediate size calcifications reduced it by 9% and 5%, respectively, when placed midway between the source and thedetector. Unlike currently prevalent low- and medium-Z scintillators, high-Z crystals are sensitive to characterization and in vivo measurement conditions. However, good agreement between MC data for ZnSe in the present study and experimental data for ZnSe:O by Jørgensenetal. (2021) suggests that detector signal is proportional to the average absorbed dose to the detector cavity. This enables an easy correction for non-TG43-like scenarios (e.g., patient sizes and calcifications) through MC simulations. Such information should be provided to the clinic by the detector vendors.

A rapid and direct method for half value layer calculations for nuclear safety studies using MCNPX Monte Carlo code

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.

Molecular Polar Surface Area, Total Solvent Accessible Surface Area (SASA), Heat of Formation, and Gamma-Ray Attenuation Properties of Some Flavonoids

The chemical and physical characteristics of several flavonoid compounds such as geraniol, thymoquinone, betaine, apigenin, N-acetylcysteine, catechin, l-carnosine, epigallocatachin, and saponarin were examined in this work. Numerous molecular properties of all flavonoid compounds used in this study were calculated using the Calculate Molecular Properties module of Accelrys Discovery Studio v20.1.0.19295.0. These properties included molecular polar surface area, total solvent accessible surface area, and heat of formation. We used the MCNPX general-purpose Monte Carlo code in combination with the Phy-X PSD software to determine gamma-ray interaction parameters such as attenuation coefficients, effective atomic numbers, and buildup factors. The findings indicate that the flavonoids’ elemental compositions have a direct effect on their chemical and physical properties. Additionally, a synergistic interaction of chemical and physical behaviors has been observed. Among the flavonoids studied, saporanin was shown to have the highest polar surface area and solvent accessible surface area, as well as the highest stability. Additionally, saporanin had the strongest gamma-ray attenuation characteristics across a broad photon energy range. It may be inferred that saporanin’s elemental structure enables a synergistic relationship between its chemical and physical characteristics. The findings of this study may contribute to the evaluation of saporanin’s hypoglycemic, antibacterial, and hepatoprotective effects.

Transmission Factor (TF) Behavior of Bi2O3–TeO2–Na2O–TiO2–ZnO Glass System: A Monte Carlo Simulation Study

The main objective of the present work was to assess the gamma radiation shielding competencies and gamma radiation transmission factors (TFs) for some tellurite glasses in the form of Bi2O3–TeO2–Na2O–TiO2–ZnO. MCNPX general-purpose Monte Carlo code (version 2.6.0) was utilized for the determination of TF values at various well-known radioisotope energies for different glass thicknesses from 0.5 cm to 3 cm. Moreover, some important gamma ray shielding properties were also determined in the 0.015–15 MeV energy range. The results show that glass densities were improved from 5.401 g/cm3 to 6.138 g·cm3 as a function of Bi2O3 increment in the glass composition. A S5 glass sample with the maximum Bi2O3 additive was reported with superior gamma ray shielding properties among the studied glasses. It can be concluded that Bi2O3 can be used as a functional tool in terms of improving glass density and, accordingly, gamma ray shielding attenuation properties of tellurite glasses, where the role Bi2O3 is also critical for other material properties, such as structural, optical, and mechanical.