Enhancing Precision in Radiotherapy Delivery: Validating Monte Carlo Simulation Models for 6 MV Elekta Synergy Agility LINAC Photon Beam Using Two Models of the GAMOS Code

The most crucial requirement in radiation therapy treatment planning is a fast and accurate treatment planning system that minimizes damage to healthy tissues surrounding cancer cells. The use of Monte Carlo toolkits has become indispensable for research aimed at precisely determining the dose in radiotherapy. Among the numerous algorithms developed in recent years, the GAMOS code, which utilizes the Geant4 toolkit for Monte Carlo simula-tions, incorporates various electromagnetic physics models and multiple scattering models for simulating particle interactions with matter. This makes it a valuable tool for dose calculations in medical applications and throughout the patient’s volume. The aim of this present work aims to vali-date the GAMOS code for the simulation of a 6 MV photon-beam output from the Elekta Synergy Agility linear accelerator. The simulation involves mod-eling the major components of the accelerator head and the interactions of the radiation beam with a homogeneous water phantom and particle information was collected following the modeling of the phase space. This space was po-sitioned under the X and Y jaws, utilizing three electromagnetic physics mod-els of the GAMOS code: Standard, Penelope, and Low-Energy, along with three multiple scattering models: Goudsmit-Saunderson, Urban, and Wentzel-VI. The obtained phase space file was used as a particle source to simulate dose distributions (depth-dose and dose profile) for field sizes of 5 × 5 cm2 and 10 × 10 cm2 at depths of 10 cm and 20 cm in a water phantom, with a source-surface distance (SSD) of 90 cm from the target. We compared the three electromagnetic physics models and the three multiple scattering mod-els of the GAMOS code to experimental results. Validation of our results was performed using the gamma index, with an acceptability criterion of 3% for the dose difference (DD) and 3 mm for the distance-to-agreement (DTA). We achieved agreements of 94% and 96%, respectively, between simulation and experimentation for the three electromagnetic physics models and three mul-tiple scattering models, for field sizes of 5 × 5 cm2 and 10 × 10 cm2 for depth-dose curves. For dose profile curves, a good agreement of 100% was found between simulation and experimentation for the three electromagnetic physics models, as well as for the three multiple scattering models for a field size of 5 × 5 cm2 at 10 cm and 20 cm depths. For a field size of 10 × 10 cm2, the Penelope model dominated with 98% for 10 cm, along with the three multiple scattering models. The Penelope model and the Standard model, along with the three multiple scattering models, dominated with 100% for 20 cm. Our study, which compared these different GAMOS code models, can be crucial for enhancing the accuracy and quality of radiotherapy, contributing to more effective patient treatment. Our research compares various electro-magnetic physics models and multiple scattering models with experimental measurements, enabling us to choose the models that produce the most reli-able results, thereby directly impacting the quality of simulations. This en-hances confidence in using these models for treatment planning. Our re-search consistently contributes to the progress of Monte Carlo simulation techniques in radiation therapy, enriching the scientific literature.

Assessment of Radiological Hazards of Sedimentary, Igneous and Sediments Natural Rocks

Gamma-ray spectroscopy based on a 100% efficiency hyper-pure germanium detector was used to evaluate the activity concentrations of 226Ra, 232Th, and 40K natural radionuclides in sedimentary, conglomerate, igneous and sedi-ments rock samples collected from four different locations in Eastern desert in Egypt. The obtained activity concentrations are used to evaluate the radi-ological hazards indices, absorbed dose rate, annual effective dose equivalent in air, radium equivalent, external and internal hazard index, radiation level index, annual gonadal dose equivalent, excess lifetime cancer risk and expo-sure rate. The results show that 1) the absorbed dose rate depends on the rock type, 2) the annual effective dose equivalent in air in 71% of sample below 20 mSvy-1 (permissible limit for workers), 3) the conglomerate rocks show low radioactivity level, 4) sedimentary rocks are rich in radium while igneous rocks are rich in thorium and the sediments rocks are rich in both radium and thorium.

Proposal of a Deuterium-Deuterium Fusion Reactor Intended for a Large Power Plant

This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is considered for a power plant. However, as shown in this article, even if a D-D reactor would be necessarily much bigger than a D-T reactor due to the much weaker fusion reactivity of the D-D fusion compared to the D-T fusion, a D-D reactor size would remain under an acceptable size. Indeed, a D-D power plant would be necessarily large and powerful, i.e. the net electric power would be equal to a minimum of 1.2 GWe and preferably above 10 GWe. A D-D reactor would be less complex than a D-T reactor as it is not necessary to obtain Tritium from the reactor itself. It is proposed the same type of reactor yet proposed by the author in a previous article, i.e. a Stellarator “racetrack” magnetic loop. The working of this reactor is continuous. It is reminded that the Deuterium is relatively abundant on the sea water, and so it constitutes an almost inexhaustible source of energy. Thanks to secondary fusions (D-T and D-He3) which both occur at an appreciable level above 100 keV, plasma can stabilize around such high equilibrium energy (i.e. between 100 and 150 keV). The mechanical gain (Q) of such reactor increases with the internal pipe radius, up to 4.5 m. A radius of 4.5 m permits a mechanical gain (Q) of about 17 which thanks to a modern thermo-dynamical conversion would lead to convert about 21% of the thermal power issued from the D-D reactor in a net electric power of 20 GWe. The goal of the article is to create a physical model of the D-D reactor so as to estimate this one without the need of a simulator and finally to estimate the dimensions, power and yield of such D-D reactor for different net electrical powers. The difficulties of the modeling of such reactor are listed in this article and would certainly be applicable to a future D-He3 reactor, if any.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

Increasing the Efficiency and Level of Environmental Safety of Pro-Environmental City Heat Supply Technologies by Low Power Nuclear Plants

In connection with the current prospect of decarbonization of coal energy through the use of small nuclear power plants (SNPPs) at existing TPPs as heat sources for heat supply to municipal heating networks, there is a technological need to improve heat supply schemes to increase their environmental friendliness and efficiency. The paper proves the feasibility of using the heat-feeding mode of ASHPs for urban heat supply by heating the network water with steam taken from the turbine. The ratio of electric and thermal power of a “nuclear” combined heat and power plant is given. The advantage of using a heat pump, which provides twice as much electrical power with the same heat output, is established. Taking into account that heat in these modes is supplied with different potential, the energy efficiency was used to compare these options. To increase the heat supply capacity, a scheme with the use of a high-pressure heater in the backpressure mode and with the heating of network water with hot steam was proposed. Heat supply from ASHPs is efficient and environmentally friendly even in the case of significant remoteness of heat consumers.

Determination of Timer Error and Evaluation of Its Effect on Dose for OB6, GammaBeam X200 and X-Ray Irradiators at the Secondary Standard Dosimetry Laboratory in Nigeria

Timer error as well as its convention is very important for dose accuracy during irradiation. This paper determines the timer error of irradiators at Secondary Standard Dosimetry Laboratory (SSDL) in Nigeria. The irradiators are Cs-137 OB6 irradiator and X-ray irradiators at the Protection level SSDL;and Co-60 irradiator at the Therapy Level SSDL. PTW UNIDOS electrometer and LS01 Ionization chamber were used at the Protection Level to obtain doses for both Cs-137 OB6 and X-ray irradiators while an IBA farmer type ionization chamber and an IBA DOSE 1 electrometer were used at the Protection Level SSDL. Single/multiple exposure method and graphical method were used in the determination of the timer error for the three irradiators. The timer error obtained for Cs-137 OB6 irradiator was 0.48 ± 0.01 s, the timer error for the X-ray irradiator was 0.09 ± 0.01 s while the timer error obtained for GammaBeam X200 was 1.21 ± 0.04 s. It was observed that the timer error is not affected by source to detector distance. It was also observed that the timer error of Co-60 Gamma X200 irradiator is increasing with the age of the machine. Source to detector distance and field size do not contribute towards the timer error of the irradiators. The timer error of the Co-60 Gamma X200 irradiator (the only irradiator among the irradiators with a pneumatic system) increases with the age of the irradiator.

Effect of Ultraviolet Radiation on Hsp70 Protein Expression in HaCaT Cells

Ultraviolet radiation by its wavelength is divided into: UVA, UVB and UVC. Only UVA and UVB manage to penetrate the ozone layer, but due to anthropological activities, all of them are capable of interacting with humans to a greater or lesser extent, and can generate adverse effects such as cellular stress when interacting with intra-and extracellular biomolecules. The skin is the first organ in contact with UV radiation, and the stress it generates can be analyzed by the expression of a bioindicator of cellular damage such as Hsp70. Therefore, the objective of the project was: to determine the effect of UVA, UVB and UVC radiation on HaCaT epithelial cells, by analyzing the expression of Hsp70. Materials and methods: HaCaT cells were cultured in vitro, which were irradiated with UVA, UVB and UVC light at different doses, to subsequently determine the degree of Hsp70 expression by Immunodetection by PAGE-SDS and Western Blot. Results: Basal expression of Hsp70 was observed in no irradiated HaCaT cells. When HaCaT cells were irradiated with UVA, UVB, UVC, an increase in this Hsp70 protein was observed. With UVA, a higher degree of expression was observed at a time of 30 minutes of irradiation. With UVB the highest expression shifted to a time of 20 minutes. With UVC, overexpression was observed after 10 minutes. Conclusion: UV radiation generates cellular stress on HaCaT cells, evaluated by the stress bioindicator Hsp70. According to the wavelength of UV radiation, those that have a shorter wavelength have a greater potential for cellular damage, such as UVC.

Multi-Elemental Analysis and 2D Image Mapping within Roots, Leaves and Seeds from O. glaberrima Rice Plants Using Micro-PIXE Technique

Understanding metal accumulation at organ level in roots, leaves and seeds in O. glaberrima (OG) is crucial for improving physiological and metabolic aspects in growing Asian and African rice in salted areas. The micro-analytical imaging techniques are required to reveal its accumulation and distribution within plant tissues. PIXE studies have been performed to determine different elements in rice plants. The existing microbeam analytical technique at the iThemba LABS will be applied for the 2D image mapping of fresh rice tissues to perform a concentration of low atomic mass elements (such as Al, Si, P, S, Cl, Ca, Ti, Mn, Fe, Cu, Br, Zn and K) with detection limits of typically 1-10 μg/g. Comparison of the distribution of the elements between leaves, root and seed samples using uptake and distribution of elements in particular environmental conditions with potential amount of salt in water have been performed. We are also expecting to indicate metal exclusion as salt tolerance strategies from leaves, root, and seed compartments using matrix correlation between samples and between elements on rice species.

Gross Alpha and Beta Activities and Related Lifetime Risks Assessment Due to Ingestion of Drinking Water from Different Sources in the District of Abidjan, Cote d’Ivoire

Drinking good quality water is essential for better health. It is therefore essential to assess the radiological quality of all water consumed in the District of Abidjan in order to prevent related hazards. Thus, the objective of this study was to assess the risk of cancer due to the ingestion of alpha and beta emitting radionuclides in the different types of water consumed in the region. A total of 63 water samples with 43 tap water samples, 5 bottled mineral water and 15 sachet water samples was collected and taken to GAEC laboratory for analysis. The low background Gas-less Automatic Alpha/Beta counting system (Canberra iMaticTM) was used to determine alpha and beta activity concentrations. Activity concentrations of both gross alpha and gross beta obtained in water sample were respectively lower than the WHO recommended limits of 0.1 Bq/l and 1 Bq/l. Also, the annual effective dose and total equivalent effective dose found in mineral bottled water samples were higher than in other types of water. The assessment of radiological lifetime risk has shown values of cancer risk due to ingestion alpha and beta emitters lower than recommended limit. These results indicate that there is no health hazard associated to consumption of water in the District of Abidjan.

Numerical Analysis of Heating Technique in Corium Melt Pool Convection Flow Field &Thermal Interaction in a Volumetrically Heated Molten Pool

In-Vessel Retention (IVR) is one of the existing strategies of severe accident management of LWR, which intends to stabilize and isolate corium & fission products inside the reactor pressure vessel (RPV) and primary containment structure. Since it has become an important safety objective for nuclear reactors, it is therefore needed to model and evaluate relevant phenomena of IVR strategy in assessing safety of nuclear power reactors. One of the relevant phenomena during accident progression in the oxidic pool is non-uniform high heat generation occurring at large scale. Consequently, direct experimental studies at these scales are not possible. The role computer codes and models are therefore important in order to transpose experimental results to reactor safety applications. In this paper, the state-of-the-art ANSYS FLUENT CFD code is used to simulate Non-uniform heat generation in the lower plenum by the application of Cartridge heating under severe accident conditions to derive the basic accident scenario. However, very few studies have been performed to simulate non-uniform decay heat generation by Cartridge heaters in a pool corresponding lower plenum of power reactor. The current investigation focuses on non-uniform heating in the fluid domain by Cartridge heaters, which has been done using ANSYS FLUENT CFD code by K-epsilon model. The computed results are based on qualitative assessment in the form of temperature and velocity contour and quantitative assessment in terms of temperature and heat flux distribution to assess the impact of heating method on natural convective fluid flow and heat transfer.

System Variables Design of Safety Analysis for Fast Reactors

This research aims to examine the risk in the technology design of fast breeder reactors while the development depends on safety considerations. The project explored the variables, which could affect positively the expected average fuel burn-up, breeding ratio, and decay heat removal. That is accomplished using features such as guard vessels and elevated pipe routing to prevent the cracked state of both core components and fuel cladding interface conditions. So, the cracked region of fuel was detected by thermal-hydraulic analysis. We used ZrFeCr alloys to estimating of the rise in fuel cladding and coolant that can be incorporated in the design ZrFeCr alloys to uniform corrosion in temperature and 10.3 Mpa pressure. Fast creep of the reactor vessel during the coolant heat-up transient is another issue to be considered corrosion resistance of structural material can be achieved by controlling oxygen content in steel alloy. In this trend, S4337 S5140 steels are wide and can be used in future fossil power plants because of their excellent high-temperature strength.

Feasibility to Convert the NuScale SMR from UO2 to a Mixed (U, Th)O2 Core: A Parametric Study of Fuel Element—Seed-Blanket Concept

Delays in the construction of nuclear reactors due to licensing issues have been a problem across the world, affecting projects in Finland, France, and the United States. Small Modular Reactors (SMRs) emerge as a transition between Generations III+ and IV in order to make nuclear energy more competitive with other energy sources, including renewables. In this study, the SMR NuScale, one of the most promising projects today, is investigated for its conversion into a U-233-producing reactor through the Radkowsky seed-blanket fuel element concept, applied in the Shippingport reactor, in a parametric study. Initially, a validation of the reference reactor (NuScale) was carried out with data from technical documents and papers, thus demonstrating the agreement of the computational model carried out with the SERPENT code. Then, a parametric study is carried out to define the area of the seed and blanket region, proportions of enrichment and pitch length. Finally, a comparison is made between the production of U-233, TRU reduction, burn-up extension and neutronic and thermohydraulic safety parameters. This study demonstrates an improvement in the conversion factor and a considerable reduction in the production of TRU, in addition to the production of U-233 with a low proportion of other uranium isotopes that can lead to the beginning of the thorium cycle with already consolidated technologies.

Low-Dose Gamma Radiation Fields Decrease Cell Viability, Damage DNA, and Increase the Expression of Hsp70 and p53 Proteins in Human Leukocytes

Ionizing radiations are tools in diagnosis and treatment of diseases. Leukopenia from exposure to ionizing radiation has been reported. Due to their radiosensitivity, leukocytes are a biological model to analyze cell damage. Therefore, cell viability, DNA damage, and Hsp70 and p53 expression in human leukocytes exposed to low-dose gamma radiation fields from a 137Cs source were evaluated. A decrease in cell viability, DNA damage and an increase in the expression of Hsp70 and p53 proportional to the radiation dose received was found, which was 0.2, 0.4, 0.6, 0.8 and 1.0 mGy.

Simulation of the Traweling Wave Burning Regime on Epithermal Neutrons

New results of two computer experiments on modeling of superthermal neutron-nuclear combustion of natural uranium for two different flux densities of external neutron source and duration of half a year each are presented. The simulation results demonstrate the dependence of the autowave combustion modes on the parameters of the external source.

Evaluation of Radon Concentration and the Health Risk in the Offices of the Institute of Science and Technology of the “École Normale Supérieure”, Burkina Faso

The second most important cause of lung cancer after smoking is radon gas. Thus, the determination of indoor radon concentrations in residential buildings and workplaces is an important public health concern. The purpose of this research was to measure the concentration of radon gas in the offices of the Institute of Science and Technology and to evaluate the effective dose in the lungs and the risk of cancer. This study used Corentium’s AIR THINGS digital radon detector to determine the radon concentration in sixteen (16) offices. The digital radon detector air Things of Corentium was placed in each office for a minimum period of one week and the concentration values were recorded every 24 hours. The values recorded in each office were the short-term average and the long-term average during seven days of measurement. The short-term radon concentrations vary between 5.286 Bq/m3 and 192.714 Bq/m3 with an average of 48.01 Bq/m3 and those in the long-term were between 6.143 Bq/m3 and 172.571 Bq/m3 with an average of 52.46 Bq/m3. The measurements in office N°6 and 13 were above the lower limit of 100 Bq/m3 proposed by the WHO. The short-term and long-term effective doses in the lungs for offices N°6 and 13 were above the “normal” background level of 1.1 mSv/year proposed by UNSCEAR-2000. The short-term effective dose in the lungs for office N°6 was above the lower limit of 3 mSv per the ICRP-23 recommendation. The average number of lung cancer cases per year per million people was 15.

Development of a LDIE Prediction Theory in the Condition of Magnetite Formation on Secondary Side Piping in Nuclear Power Plants

It has been thought that wall thinning on the secondary side piping in nu-clear power plants is mostly caused by Flow-Accelerated Corrosion (FAC). Recently, it has been seen that wall thinning on the secondary side piping carrying two-phase flow is caused by not only FAC but also Liquid Droplet Impingement Erosion (LDIE). Moreover, it turns out that LDIE in nuclear power plants does not result from a single degradation mechanism but also from the simultaneous happenings of LDIE and FAC. This paper presents a comparison of the mass loss rate of the tested materials between carbon steel (A106 B) and low alloy steel (A335 P22) resulting from degradation effect. An experimental facility was set up to develop a prediction model for clarifying multiple degradation mechanisms that occur together. The experimental facility allows examining liquid droplet impingement erosion in the same conditions as the secondary side piping in nuclear power plants by generating the magnetite on the surface of the test materials. The magnetite is formed by controlling the water chemistry and the temperature of fluid inside the facility. In the initial stage of the experiments, the mass loss rate of A106 B was greater than that of A335 P22. However, after a certain period of time, the mass loss rate of A335 P22 became greater than that of A106 B. It is presumed that the results are caused by the different yield strengths of the test materials and the different degrees of buffer action of the magnetite deposited on their surfaces. The layer of magnetite on the surface of A106 is thicker than that of A335 P22, due to the different amount of chrome content. In nuclear power plants, carbon steel piping having experienced wall thinning degradation is generally replaced with low-alloy steel piping. However, the materials of pipes carrying two-phase flow should be selected considering their susceptibility to LDIE.

SVBR-100 Nuclear Technology as a Possible Option for Developing Countries

Many developing countries need ecologically clean power sources (PS). The nuclear power plants are such sources. However, a great number of the developing countries do not possess developed large capacity power systems. Moreover, currently in the developing countries, there are no highly skilled personnel to provide construction and reliable and safe operation of the nuclear plants, which are complex and potentially hazardous systems. In some countries, the level of terroristic threat is extremely high. For that reason, there are specific requirements to the nuclear PSs intended for use in the developing countries. In the presented report, the specific requirements which must be met by the NPT proposed for use in developing countries are formulated, basic statements of the SVBR-100 concept are presented, design and principal scheme of the reactor fa-ility are described, major characteristics of SVBR-100 are summarized.

Adsorption Characterization of Strontium on PAN/Zeolite Composite Adsorbent

This work reports the adsorption of strontium from aqueous solutions onto PAN/zeolite composite. The strontium adsorption on the composite adsorbent was studied as a function of initial strontium concentration, pH of the solution, contact time and temperature. Adsorption isotherms like Langmuir, Freundlich,Dubinin–Radushkevich (D–R) and Temkin were used to analyze the equilibrium data at the different concentrations.Adsorption process well fitted to Temkin isotherm model. Thermodynamic parameters such as the changes in enthalpy, entropy and Gibbs’ free energy were determined, showing adsorption to be an exothermic and spontaneous process.

The Diffusion of Tc-99 in Beishan Granite-Temperature Effect

In the safety assessment of a potential site for high-level radioactive wastes (HLW) disposal, the investigation on the geochemical behaviors of key radionuclides with the possibility for releasing from the potential repository is an important aspect. Due to the high mobility of technetium under most repository conditions, lots of research works were performed to investigate the diffusion of technetium in different potential rocks. In spite of these studies, there remains a lack of data addressing temperature effects. In this paper, the diffusion of 99Tc in Beishan granite at temperatures from 25℃ to 55℃ was studied with laboratory small scale diffusion devices. The experimental data were fitted with a finite difference scheme to get the effective diffusion coefficient (De) of . The results indicated that the relationship of De with temperatures could be described as the modified Stokes-Einstein equation, and the formation factor of Beishan granite was constant in the temperature range of 25℃ - 55℃ with the value of (3.91 ± 1.77) × 10-4.

Development of ToSPACE for Pipe Wall Thinning Management in Nuclear Power Plants

A number of piping components in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), cavitation, flashing, SPE (Solid Particle Erosion), LDIE (Liquid Droplet Impingement Erosion), etc. Those mechanisms may lead to thinning, leak, or rupture of the components. Due to the pipe ruptures caused by wall thinning in Surry unit 2 of USA in 1986 and in Mihama unit 3 of Japan in 1994, the pipe wall thinning management has emerged as one of the most important issues in nuclear power plants. To manage the pipe wall thinning in the secondary system, Korea has used a foreign program since 1996. As using the foreign country’s program for long term, it was necessary to improve from the perspective of the users. Accordingly, KEPCO-E & C has started to develop the 3D-based pipe wall thinning management program (ToSPACE, Total Solution for Piping And Component Engineering management) from eight years ago, and the development was successful. This paper describes the major functions included in ToSPACE program, such as 3D-based DB (Database) buildup, development of FAC and erosion evaluation theories, UT (Ultra-sonic Test) data reliability analysis, field connection with 3D, automatic establishment of long-term inspection plan, etc. ToSPACE program was developed to allow site engineers performing the selection of inspection quantity at each refueling outage, UT data reliability analysis, UT evaluation, determination of next inspection timing, identification of the inspecting and replacing components in 3D drawings, etc., to access easily.