As the amount of power generated by nuclear power plants increases, the importance of managing Spent Nuclear Fuel (SNF) to prevent proliferation increases as well. In a previous study, we optimally designed the gamma emission tomography instrument named Yonsei Single-photon Emission Computed Tomography version 2 (YSECT.v.2) based on Monte Carlo (MC) simulation for inspection and detection of defects in SNF. The objective of the current study was to fabricate a prototype YSECT.v.2 instrument and to evaluate its performance. The YSECT.v.2 instrument consists of scintillator-based detection modules, a multi-channel data acquisition (DAQ) module, a heat reduction module, and a waterproof housing. Based on the experimental results for a mock-up of fresh nuclear fuel rods, the spatial resolution of the prototype YSECT.v.2 instrument was determined to be 10.8 mm. In the experiment with ¹³⁷Cs test sources, a high-quality image could be obtained with the exclusion of the pixel value below the threshold. For a 6 × 6 array, the spatial resolution for ¹³⁷Cs was analyzed to be 7.2 mm. Considering these results, we expected that source distribution could be distinguishable using the YSECT.v.2. In the future, further experimentation will validate the performance of the fabricated instrument for a mock-up of SNF rods in water.

As the use of nuclear and radiation technologies increases, the importance of radiation safety and monitoring increases. In this study, we develop a nuclear facility monitoring system (NFMS) for rapid response to radiation accidents in nuclear material storage facilities; (1) multi-sensor network based on NaI(Tl) detector and FPGA-DAQ system and (2) an artificial neural network (ANN) algorithm for tracking radiation source location. Energy resolution and sensitivity of the detectors were evaluated to accurately track the location of radioactive materials and to identify nuclides using the multi-sensor network. To evaluate the localization accuracy of NFMS, a test storage facility was built and experiments were performed. Localization accuracy was obtained by analyzing the measured counts for each detector using an artificial intelligence (AI) based ANN algorithm, confirming an accuracy of over 99 %. The developed NFMS is expected to contribute to the safe management of radioactive materials in nuclear facilities.

Bonding between plasma-facing materials (PFMs) and heat sinks is an important factor for the performance of plasma-facing components (PFCs) in the tokamak device; in some specific applications, graphite is adopted as the plasma-facing material and CuCrZr as the heatsink. Bonding of graphite and CuCrZr is a critical and difficult process due to the poor wettability of graphite and wide gap of material properties between graphite and CuCrZr, to which should be paid more attention. Herein, brazing processes were developed to achieve the bonding of CuCrZr and graphite, and a layer of oxygen-free copper (OFC) was placed between graphite tiles and CuCrZr heat sinks as a stress-buffer layer. Three types of brazing fillers including Ti powder with AgCu foil, AgCuTi paste and AgCuTi foil were used to achieve the bonding of CuCrZr and graphite. Samples were tested by scanning electron microscopy (SEM), energy dispersive spectrometry (EDS) and shear strength experiment; the results showed that these three types of fillers can achieve the bonding of CuCrZr and graphite. In addition, one-step brazing and multi-step brazing were both developed to achieve the bonding of graphite/OFC/CuCrZr, and the mock-ups were manufactured and endured cyclic high heat loads in the high heat flux (HHF) tests. In conclusion, the simplified one step brazing process including types of fillers, brazing temperature and brazing sequence was determined, which provides a promising premise method for the heat transfer structure design of PFCs.

In today's era, the green economy has become the central focus worldwide, along with the clean energy transition. Unfortunately, the current economic development model flourishes at the expense of the natural environment and ecosystems. Thereby, the concept of green economy gets immense attention among scholars and policy analyst. Most countries are dependent on energy import due to limited and unsustainable natural resource use. In this context, environmental taxation can play a key role in restricting the unsustainable use of natural resources, and nuclear energy can reduce energy import dependency, helping to shift from conventional growth to green economic growth. In this context, this study is an attempt to explore the relationship between nuclear energy, environmental taxation, greenfield investment, and the green economy in a panel of 16 Organization for Economic Co-operation and Development (OECD) economies from 2000 to 2019. To get empirical estimate the study relies on Method of Moment Quantile Regression (MM-QR) approach. Interestingly, this study found that nuclear energy development discourages green economic growth across all quantiles in OECD countries. Furthermore, the study found that environmental taxation helps OECD countries achieve green economy status. Greenfield investments are another factor promotes the green economy. Climate policy measures are suggested in line with increasing share of nuclear energy in energy mix through greenfield investment in energy sector.

This paper presents the development process of a high-dose X-band linear accelerator (LINAC) for radiotherapy machines, specifically focusing on stereotactic body radiation therapy (SBRT). The development process includes a multi-physics design, approach for the radio frequency cavity in the LINAC, with an emphasis on minimizing beam size and maximizing shunt impedance of the LINAC. Commissioning tests were conducted to evaluate the performance of the LINAC, which involved measurements of beam energy, beam size, beam current, and X-ray dose rate.

The permeability of the ore-bearing layer is an important indicator affecting the in-situ leaching (ISL) of uranium-bearing sandstone, which is related to various factors such as pore shape, distribution, and size. In order to study the effect of pore structure on seepage in low-permeability uranium-bearing sandstone, CT scanning tests were conducted to create a 3D digital core based on scanning images and to calculate the fractal dimension using the box counting dimension method, which integrated fractal theory to define the core samples' pore structure. The permeability prediction was realized based on the porosity-permeability model and the fractal theory model. Results indicated that this type of sandstone is obviously characterized by pore connectivity, large differences in distribution, and strong microscopic inhomogeneity. The pores are dominated by micro- and nanopores, as well as small pores, accounting for 90 %; macropores are few in number, but the diameters of their single pores are large. The distribution of pore structure in this type of sandstone exhibits a good fractal characteristic; the three-dimensional fractal dimensionality is 2.044-2.310. The porosity-permeability model was established, and permeability prediction was realized by combining the fractal theory to provide theoretical support for determining the values of well field parameters in ISL.

The radiological and environmental hazards of Wadi El Markh granitic rocks were investigated, and magnetic data were used to delineate the structural framework and determine the bedrock depth in the area. The result displayed that geological structures influence the occurrence of uranium mineralization in this area and are mainly associated with altered granitic formations. The activity concentrations of radioisotopes in the regions studied showed a range of values: ²³⁸U varied between 345 and 1729 Bq.kg^-1, averaging 980 Bq.kg^-1; ²³²Th ranged from 73 to 162 Bq.kg^-1, with an average of 120 Bq.kg^-1; and ⁴⁰K varied from 829 to 1790 Bq.kg^-1, average 1245 Bq.kg^-1. The measured concentrations of ²³⁸U, ²³²Th, and ⁴⁰K in the analyzed granitic rocks samples exceeded the worldwide average of 35, 45, and 412 Bq.kg^-1, respectively. The primary radiological risks related to these granitic rocks were associated with the gamma radiation emitted by the radioactive elements. The statistical assessment confirmed that the main contributors to the radiological risks were uranium, potassium, and associated minerals in the granitic rocks. The entire investigation region has been determined to exceed the permissible safe radiation dose rate limit of 1 mSv/year. As a result, the study determined that the granitic rocks in the surveyed area were deemed unsuitable for construction because of their elevated levels of radioactivity. The effects of pollution on the ecological system were evaluated using several ecological indices, including the Geoaccumulation index (Igeo), Contamination factor (CF), Degree of Contamination (Cd), Pollution Load Index (PLI), Potential ecological risk factor (Eri), and Potential Ecological Risk Index (RI). Based on Cd and PLI, 100 % of the samples were found to be very highly polluted to the ecological system and suggest deterioration if used. Regarding RI, the metals were arranged as Cd > As > Co > Cu > Pb > Cr > V > Zn, with considerable risk in all samples.

Accurate measurement of the dose delivered to patients is crucial in any radiation therapy, including C ion beam radiotherapy, to ensure effective treatment and minimize the risk of damage to healthy tissues. The accessibility of thermoluminescence (TL) - based materials for carbon ion beam dosimetry continues to be constrained. With the aim of this, an investigation has been done to study the TL properties of the Terbium (Tb) doped LiCaAlF₆ phosphor following irradiation with C ion beam. The LiCaAlF₆: Tb phosphor has been synthesized using melting method in argon gas atmosphere. TL glow curves has been analyzed on exposure to carbon ion beams at energies of 85 MeV and 50 MeV within the fluence range of 10⁹ - 10¹² ions/cm². The TRIM code, relying on Monte Carlo simulation, has been employed to compute parameters such as absorbed dose, ion range, and energy loss. The Tb-doped LiCaAlF₆ phosphor exhibits a glow curve structure with distinctly resolved peaks at 130 ℃ and 325 ℃. Notably, the TL intensity has been significantly increased by 10⁵ times than that of undoped LiCaAlF₆. This enhancement is attributed to the generation of a high density of Tb-associated defects within the energy band gap. The TL enhancement has been elucidated based on charge compensation mechanism. Further, TL glow curve shape and structure remain consistent across different energies, i.e., 85 and 50 MeV C ion beams, indicating an energy-independent behavior of this phosphor. Nevertheless, the TL intensity at 85 MeV is higher compared to that at 50 MeV. Detailed dosimetry properties have been investigated for 85 MeV C ion beams within the fluence range 2×10⁹-5×10¹² ions/cm². The phosphor exhibits a sublinear dependence against dose for the studied fluence range with no discernible shift in peak position. The favorable outcomes include low fading during the initial days of storage time and batch homogeneity. TL glow curve analysis indicate the formation of trapping level in the range of 0.90-1.72eV within the band gap of the phosphor which are responsible for TL process. The phosphor's simple glow curve structure, minimal fading, consistent batch homogeneity and high stability of the glow curve suggest that the material can be explored for C ion beam dosimetry application.

Storage rings that use the hybrid multi-bend achromat (MBA) lattice concept [1] have been realized. The main characteristics of hybrid MBA as a typical lattice for a 4th generation storage ring are to adopt two large dispersion bumps in cell, and to make a -I-transformation between two dispersion bumps to cancel nonlinear effects. To increase the effectiveness of nonlinear control, analytical investigation of nonlinear phenomenon is essential. In this paper, we describe analytical investigation with a nonlinear driving term and simple model. We obtained similar results to the existing time-consuming MOGA optimization, but use a simplex optimizer with fewer nonlinear control knobs.

This paper elucidates a comprehensive derivation of the variational formulation pertaining to the static and spatial kinetics self-adjoint angular flux (SAAF) neutron transport equation. The methodology employed for discretization of the spatial variable is the finite element method, while the energy group discretization is executed via the group method, and the directional discretization is conducted using the discrete ordinates method. Analytic expressions for the discretization to variable separation under both vacuum and reflective boundary conditions are furnished. Constructed upon the MOOSE framework, a code designated as SAAFCGSN has been developed for the resolution of the SAAF neutron transport equation. The zero-order scattering matrix within this computational framework is managed through an innovative "decoupling" method, thereby enhancing the computational efficiency significantly. The functionality and robustness of the SAAFCGSN code are corroborated through meticulous evaluation involving seven distinct steady-state scenarios as well as two transient states. Empirical outcomes verify the compatibility of the SAAFCGSN code with both structured and unstructured mesh, inclusive of their amalgamations, thus facilitating maintenance and ensuring elevated computational accuracy. In addition, a performance analysis benchmarked against IAEA standards reveals that the adoption of the scattering matrix "decoupling" method propels a computational speed increase exceeding 30 %, signifying a notable advancement in calculation efficiency.

The current study has focused on evolving materials with high attenuation performance and vigorous mechanical properties. Nano hexagonal boron nitride (h-BN) has been synthesized from boric acid and urea and then heated at a low temperature. XRD, SEM, EDX-mapping images, and FTIR investigated the nano h-BN synthesized and boron oxide (B2O3) milled. The new nanocomposites based on ethylene propylene diene rubber (EPDM), different concentrations of h-BN NPs, and B₂O₃ have been prepared. The physical, mechanical, and thermal properties and neutron attenuation behavior of nanocomposites were characterized. The shielding properties were determined by measuring the fast neutrons and total gamma-ray attenuation coefficients of the ²³⁹Pu-α-⁹Be neutron source. It was seen that adding a coupling agent (maleic anhydride) was appropriate for improving interfacial adhesions between EPDM chains and the nanofiller h-BN and B₂O₃ compared to EPDM unloaded. From the results, we observe that EPDM loaded with nano h-BN and nano h-BN/B₂O₃ was noticeably boosted over that of the unloaded EPDM for tensile strength (TS), EPDM/nano h-BN/B₂O₃ had a high TS at a concentration of 3 % h-BN/10 % B₂O₃ (42 %) compared with unloaded EPDM. Moreover, EPDM/3 % h-BN and EPDM/3 % h-BN/10 % B₂O₃ had the highest thermal stability until 490 ℃ compared to unloaded EPDM is stable at 350 ℃. Finally, the maximum macroscopic effective shielding behaviors and removal cross-section are estimated by incorporating h-BN 3 % into EPDM.

Human reliability analysis (HRA) of nuclear research reactors often encounters a lack of human performance data, a challenge that is also faced by the TRIGA reactor of Thailand having no specific full method or human error database for HRA. To overcome this challenge, in 2023, HRA teams from the Korea Atomic Energy Research Institute (KAERI) and the Thailand Institute of Nuclear Technology (TINT) jointly developed an HRA framework for the TRIGA reactor. In the HRA framework development, the HRA practitioners applied KAERI's three main HRA tools, namely 1) the EMpirical data-Based crew Reliability Assessment and Cognitive Error analysis (EMBRACE) method, 2) the Human Reliability data EXtraction (HuREX) database, and 3) the TAsk COMplexity (TACOM) method. The HRA framework covers the overall process including the classification of documents and human error types as well as nominal human error extraction for estimating nominal human error probabilities. According to the results of the first use of the HRA framework on the TRIGA reactor in Thailand, the HRA framework provided an effective way to improve the procedures and systematically identify nominal human errors of actions in the emergency procedures.

Based on the lessons from the Fukushima accident, new strategies to mitigate the spread of accidents in nuclear power plants have been established that consider portable equipment. The diverse and flexible coping strategies (FLEX) and multi-barrier accident coping strategy (MACST) based on the utilization of portable equipment are being adopted in case of an extended loss of all AC power and loss of ultimate heat sink. In this context, the establishment of specific strategies for utilizing portable equipment has emerged as an important issue. This study proposes a multi-unit probabilistic safety assessment (MUPSA)-based evaluation framework consisting of several steps to quantitatively assess the reduction in risk from the use of portable equipment at a multi-unit site, which can provide a basis for establishing optimal portable equipment strategies. According to the framework, after first defining the scope of the evaluation, risk-significant types of portable equipment are selected and the scenarios are analyzed. Risk reduction is then evaluated using modified MUPSA models according to the deployment of the portable equipment. A case study was conducted for an internal loss of off-site power (LOOP) and seismic LOOP focusing on how to evaluate the effectiveness of portable equipment through the developed framework.

The leakage rate of aerosol particles is one of the most important parameters for source term evaluation of containment in severe accidents. In this study, gas leakage characteristics experiments were conducted on capillaries with different inner diameters and tube lengths under different pressure differential conditions. The effects of upstream and downstream pressure differential, capillary tube diameter, tube length, aerosol particle size, whether condensation or not, and release time on aerosol leakage rate were studied. The results show that when the pressure difference between the two ends of the capillary is greater than 0.2 MPa, the leakage rate of the capillary shows a pattern of first increasing and then decreasing with the increase of the pipe diameter; The influence of aerosol particle size on leakage rate mainly affects the inertial deposition effect, which increases with the increase of particle size, leading to a decrease in leakage rate; The condensation effect of steam in capillaries can lead to enhanced effects such as condensate capture, thermophoresis, and diffusion electrophoresis, which have a positive strengthening effect on the retention of aerosol particles.

This study focuses on the development of a curated separation system aimed at separating thorium ions as part of a recovery plan for managing the thorium radionuclide in the rare earth element (REE) processing industry. As a step towards advancing technology readiness, the separation system employs the electrosorption technique, which involves the migration and storage of thorium ions in the electrical double layers on the porous surface of a carbon-based electrode. Using a prototype electrosorption cell operated at an applied voltage of 1.0 V, the system successfully achieved a thorium ion recovery rate of 84.5 ± 0.29 %, with an impressive electrosorption capacity of 105.26 mg-Th/1 g-carbon. Notably, despite the presence of higher concentrations of REE as competing ions, greater selectivity towards thorium was observed which likely attributable to its larger ionic radius, higher electron affinity, and greater distribution coefficient. These findings highlight the efficacy of the curated separation system and its potential to balance between REE economic significance as well as a protection to the environment.

We introduce a novel fast-reciprocating probe head installed on the EAST (Experiment Advanced Superconducting Tokamak) containing 12 graphite tips. This allows us to measure a variety of plasma parameters such as the plasma density, temperature, and space potential in the scrape-off layer (SOL). We will also acquire the data at high frequency thus measuring the ion saturation current, I_sat, and the floating potential, V_f, average and fluctuating values. Moreover, with poloidally separated tips, we measure the poloidal electric field, E_θ, at two toroidal locations, assuming a plasma flow dominated by the E×B drift this leads to the determination of the radial velocity. As a consequence, the turbulent radial flux will thus be determined. In the parallel direction, we have four tips that are aligned parallel with the magnetic field lines, thus making an angle of 4°24' to the horizontal direction. This unique design should allow the measurement of the toroidal wavenumber, k_φ, a quantity rarely obtained in tokamaks. This new probe, with 7 tips for measuring the ion saturation current and 4 tips to measure the floating potentials, was used on the EAST during the 2023 summer campaign. We present the first experimental results obtained during an L-mode plasma.

Nuclear power plant (NPP) structures face significant challenges in ensuring structural resilience subjected to earthquake motions. The inerter-based vibration absorbers (IVAs) have been proven to control structural vibration, but they have not been applied in small modular reactor (SMR) structures. In this study, the optimal IVA parameters are investigated for SMR structures, which focus on reducing the seismic response of the first floor and fifth floor. By using a genetic algorithm (GA), the IVAs parameters are optimized to minimize the storey displacement variance and storey acceleration variance. Besides, the robustness analysis is performed to ensure the optimized parameters are practical. The various ground motions are selected for time history analysis (THA) to illustrate the seismic response of SMR equipped with the optimal IVAs. The results of THA demonstrate that the optimal IVAs enhance the seismic performance of the SMR compared with traditional SMR structures. By positioning optimal IVAs, significant mitigation of both storey displacement and storey acceleration are achieved. This study explores to a deeper understanding of IVAs' efficacy in SMR structures, which is beneficial to the practical use of IVAs in SMR structures.

Industrial processes have the potential to release cobalt into the environment, and exposure to cobalt in quantities higher than specific amounts can have harmful impacts on living organisms. This study aimed to determine whether natural dolomite from the Thrace region of Türkiye could effectively remove Co(II) from aqueous solutions. Batch studies were conducted to examine the impact of various experimental factors, including adsorbent dosage, initial Co(II) concentration, contact time, solution pH, and temperature. The cobalt concentrations in aqueous solutions were determined by employing ⁶⁰Co radionuclide as a radiotracer. Experimental results revealed that optimum adsorption conditions were obtained at a solid-liquid ratio of 20 g·L^-1, initial Co(II) concentration of 128 mg·L^-1, contact time of 120 min, and pH 6 with a maximum adsorption capacity of 2.32 mg·g^-1. The pseudo-second-order kinetic and Redlich-Peterson isotherm models described the experimental data well. The adsorption process of Co(II) ions on the dolomite was spontaneous and exothermic. The results showed the potential use of dolomite to remove such a toxic metal ion from industrial wastewater.

The radiation-induced volumetric expansion and mechanical properties of different types of irradiated aggregates are affected to varying degrees, and the effects of mineral composition of irradiated aggregates on the volume expansion and elastic modulus are not completely clarified. In this study, the prediction models for the crack volume and effective elastic moduli of irradiated aggregates were established based on the model of the polycrystalline assemblage of minerals and the self-consistent scheme, and the influence of randomness of mineral composition on the volume expansion and effective elastic moduli of irradiated aggregates was further analyzed. The predictions of the RIVE and effective elastic moduli of irradiated aggregates were in good agreement with the test reactor data in the references. The results show that the randomness of volume expansion and effective elastic moduli of irradiated aggregates is dominated by the uncertainty in the crack volume of irradiated aggregates, and it can be estimated for a specific aggregate based on the variability of the quartz content. The findings provide a foundation for predicting the probability-based degradation of mechanical properties of irradiated concrete based on the mineral composition of aggregates.

Due to its low background, alpha spectrometry is capable of determining very low activity concentrations of alpha emitters in all types of samples, including environmental samples. This low background can be lost due to detector contamination caused by recoil nuclei, which can even render the detector useless. In this paper, the thickness of a thin film used to cover the source is calculated analytically. This film, capable of retaining these nuclei and avoiding this contamination, allows all the alpha particles emitted by the source to reach the detector. The energy loss of the alpha particles passing through it was also estimated in order to identify the peaks in the spectrum, including the unexpected ones. The validity of the calculations carried out and the effectiveness of the film in retaining the recoil nuclei were verified using thorium samples. Finally, the effect of applying a negative potential to the source on the retention of these nuclei was studied. The study was carried out using a PIPS detector under vacuum conditions (0.1 mm Hg) with the samples placed at a distance of 5 mm from the detector. The recoil suppression systems of some commonly used commercial instruments were also tested.

This study investigated using rare earth elements as tracers to study wildfire impacts. The study area was Antalya, located in the Mediterranean region of Türkiye. This region is known for the frequent occurrence of wildfires, which cause significant damage to the flora and fauna of the area. A series of wildfires occurred in the region in 2016. Soil samples were collected from the affected areas to understand these wildfires' impact better. Instrumental neutron activation analysis was used to analyze the samples. The study found that the north-facing aspect had higher concentrations of rare earth elements. This could be because north-facing aspects are cooler and more humid, with less erosion, leading to better ecological restoration and less transportation of rare earth elements.

This study analyzes the single-event transient (SET) characteristics of alpha particles on multi-channel Forksheet-FET and Nanosheet-FET at the device and circuit levels through 3D TCAD simulations. The study investigates the differences in SET responses based on the energy and incident position of incoming alpha particles, considering the structural variances between Forksheet-FET and Nanosheet-FET, as well as the presence or absence of bottom dielectric isolation (BDI) in the fabrication process. Specifically, the introduction of BDI is observed to significantly suppress the voltage drop caused by 'unintended' current, as it can block the substantial electron-hole pairs (EHP) generated by injected alpha particles in the bulk substrate from reaching the FET terminals. Furthermore, it was confirmed that the size of abnormal current decreases as the energy of the injected alpha particle increases. Additionally, evaluating the response to SET based on the fundamental logic circuit, the CMOS inverter, revealed relatively small abnormal voltage drops for both Forksheet and Nanosheet when BDI was applied, confirming high immunity to radiation effects. Moreover, it can be observed that the application of BDI enhances reliability from a memory perspective by effectively suppressing voltage flips in the SRAM's cross-coupled latch circuit.

Accurate nuclide identification in γ-spectrum analysis of low- and intermediate-level radioactive waste with high-purity germanium detectors necessitates initial forced fitting with a nuclide library, yet inaccuracies in library data may lead to misidentification and missing nuclides. To this end, background clipping strategies were hereby analyzed, and a novel deconvolution spectrum analysis method was proposed, which utilized continuous wavelet transform for peak searching and Gaussian first-order derivative quadratic convolution for calculating peak width. Furthermore, to effectively realize the nuclide identification and peak area calculation, a response filter function model was established through the peak shape calibration. By eliminating the need for nuclide library parameter settings prior to overlapping peak separation, the issue of inaccurate matching arising from reliance on the precision of the nuclide library was addressed. Moreover, spectrum analysis experiments were carried out on standard point sources and 200 L drums, and the results were compared and analyzed using GammaVision. Compared to the GammaVision results set by the accurate nuclide library, the area error of strong peaks decreased from 27.5 % to 4.82 %, while that of weak peaks witnessed a decline from 49.98 % to 27.5 %. Finally, the accuracy of the proposed method was verified using the Pakistan Nuclear Library.

The high-temperature material properties of Type 316L stainless steel (hereafter referred to as "316L SS") were determined through a series of material tests to enable the design of pressure vessels and piping operating up to 700 ℃. Currently, the only applicable design rule for 316L SS components in the high-temperature creep range is French high-temperature design standard, RCC-MRx. However, the material properties provided by RCC-MRx are limited to approximately 600 ℃. In this study, new material properties and relevant design coefficients for Type 316L components subjected to 700 ℃ coolant were determined based on material tests including tension, fatigue and creep tests conducted on 316L SS specimens. Utilizing these supplemented properties and design coefficients, the design of pressure boundary components and piping made of 316L SS in a large-scale test facility known as TESET, subjected to 700 ℃ coolant, was carried out. Several large-scale sodium tests at high temperatures up to 700 ℃ were successfully conducted at the TESET facility, with the main components and piping designed and constructed using 316L SS.

A new series of tellurite glass-ceramic was prepared for potential utilization in the radiation shielding field. The thermal features of parent glasses were assessed to choose the best heat treatment technique. The morphological and structural properties of glass ceramics were investigated by scanning electron microscope (SEM), X-ray diffraction (XRD), and Fourier transform infrared (FTIR). The UV absorption was utilized to calculate the optical features of glass ceramics. The shielding properties were computed based on the Phy-X software. The transformation of glasses to glass ceramics enhanced the density of all glass ceramics. XRD results exhibited an orthorhombic BaTeO₃ crystalline phase for the TMSB0 and TMSB5 samples. It is proved that the monoclinic molybdenum tellurium oxide (Mo₅TeO₁₆) crystalline phase for TMSB10, TMSB15, and TMSB20 samples. Also, the change in crystalline phase resulted from adding MoO₃ instead of TeO₂. SEM results affirmed the role of MoO₃ in improving the structure of glass ceramics. The thermal stability and weight loss of TMSB20 samples are higher than other glasses. Also, the adding MoO₃ led to a slight enhancement in the radiation shielding properties of the prepared samples. Data indicate that adding MoO₃ to the glasses enhanced the glass-ceramic samples' thermal, structural, and radiation protection properties.

Metallic alloy fuels from fast reactors will be reprocessed by a non-aqueous electrochemical technique known as electrorefining. Electrorefining of spent metal fuel results in the accumulation of heat generating fission products especially ¹³⁷Cs in the electrolyte, which is a eutectic salt mixture of 44 wt% LiCl and 56 wt% KCl. The fission products need to be removed from the salt so as to reduce the decay heat load and contamination thereby facilitating reuse of the salt and minimizing waste generation. Melt crystallization technique is a simple separation process being pursued for separation of fission products. This is a single step process which utilizes the difference in solubility of fission products in the solid and liquid phase. Crystallization involves purification of a substance from a liquid mixture by solidification of the desired component. Numerical modeling of separation of CsCl from a LiCl-KCl eutectic mixture by solidification was carried out using ANSYS Fluent (19.2). The enrichment of CsCl during solidification, its distribution in molten salt and segregation time were evaluated. Further, solidification experiments were conducted using the molten salt mixture and the predicted numerical results were validated.

Uranium (U) has been identified as an element of interest for the safety assessment of a deep geological repository for used nuclear fuel. In this study, the sorption of U(VI) was studied in Na-Ca-Cl solutions at ionic strengths = 0.1-6 mol/kgw (m) in a CO₂ free environment at pH_m (molal H⁺ concentration where pH_m = - log m_H+) = 4-9 for MX-80 bentonite, illite and Queenston shale and at pH_m = 5-9 for limestone. U(VI) sorption on MX-80 bentonite increased with pH_m from pH_m = 4 to 6, then decreased with pH_m until pH_m = 7, and then increased again with pH_m to pH_m = 9. U(VI) sorption on illite increased with pH_m reaching a maximum at pH_m = 7, and then decreased with further increases in pH_m. The sorption behavior of U(VI) on shale was similar to that of illite, but the extent of decrease in the sorption coefficient (R_d) value with pH_m was slightly more pronounced for the shale than observed for sorption on illite at pH_m > 7. U(VI) sorption on limestone increased with pH_m up to pH_m = 8 and then seemed to be constant at pH_m = 8-9. U(VI) sorption on all four solids was independent of ionic strength (0.1-6.0 m). The 2 site protolysis non-electrostatic surface complexation and cation exchange model successfully simulated the sorption of U(VI) onto MX-80 and illite, and the optimized values of surface complexation constants were estimated.

The SPHERE experiment aimed at studying the behaviour of Minor Actinide-bearing Driver Fuel (U,Pu,Am)O_2-x by comparing sphere-packed and pelletized fuels. The irradiation experiment was performed in the High Flux Reactor at Petten from August 2013 to April 2015, and was followed by post-irradiation examinations up to mid-2017. The present work consists in a new analysis of the SPHERE experiment, focusing on the pelletized fuel, by the means of both neutronics and fuel performance codes. This study is performed in the frame of the European Project PATRICIA. The adopted methodology and the main results achieved, assessed in particular against inert gas-related experimental data, are presented in the paper.

Despite their initial commercialization, CdTe-based materials are continuously being developed. However, an approach for obtaining the X-ray response with a current-mode CdMnTeSe (CMTS) detector remains to be explored. This study investigated the feasibility of using CMTS as a potential material in diagnostic X-ray detectors. The CMTS crystals were grown using the Bridgman method and treated with a radiation detector. The X-ray response of the CMTS detector was evaluated under various conditions, including tube voltage, tube current, and bias voltage. In addition, the mass attenuation coefficients of CdTe-based materials, including CMTS, were investigated. Our results demonstrated that CMTS exhibited promising properties for diagnostic X-ray detection, with stable and linear responses across various tube voltages and currents. In addition, the calculated sensitivity of the CMTS detector increased at higher bias voltages, demonstrating its practical applicability. However, problems such as signal loss were observed at higher dose rates, suggesting the need for further optimization of the CMTS growth techniques. Overall, this study indicated the requirement for further improvement in the CMTS crystal properties; however, the potential of CMTS as a candidate material for the next generation of diagnostic X-ray detectors was evident.

Evaluation of the leakage rate is necessary for infrastructure where hermetic performance holds critical importance. This study conducted a test to evaluate the air leakage rate of cracked reinforced concrete panels. Hollow beam specimens with an air chamber in the web were prepared to provide a relatively simple setup for airtight pressure testing under 200 kPa pressure. While testing, the lower flange of the beam specimen was subjected to tension cracking under transverse loading, and air pressure was applied to the air chamber of the beam to measure the air leakage rate through the flange plate as the magnitude of loading increased. The test variables were the differential air pressure, flange plate thickness, and reinforcement ratio. The test results showed that the leakage rate increased in proportion to the differential pressure. When the reinforcement ratio increased from 0.7 to 1.3 %, the leakage rate decreased by 90 %. When the plate thickness increased from 75 to 100 mm, the leakage rate decreased by 91 %. The analysis of the test results showed that the leakage rate was proportional to the cube of the crack width (R² = 0.91-0.95), which agreed with existing prediction models. When the crack width at rebar location was used, the existing models showed the best predictions for test results. This result indicates that to accurately predict the air leakage rate of reinforced concrete members, the critical crack width at the location of reinforcement (i.e. the smallest crack width in the member thickness) is more important than the crack width at the member surface.

Spent nuclear fuel (SNF) requires accurate and effective supervision to prevent radiation release to the public and the environment. Gamma emission tomography (GET) has been used to inspect partial-defects at the pin-by-pin level through the acquisition of tomographic images. However, due to the high density of nuclear fuel rods, GET shows relatively low accuracy for the central region. This study proposes an image reconstruction method for the internal fuels based on Monte Carlo simulation and machine learning algorithm. The gammas from the SNF were measured with the GET device and the sinograms were reconstructed with Maximum-Likelihood Expectation-Maximization (MLEM) with prior-aware information of SNF, and image quality was improved with the machine learning technique. Our results show that the MLEM with prior-aware information could dramatically improve the image quality and the denoising technique with the machine learning could clearly correct the over-and under-estimation. Based on the results of quantitative validation of the proposed reconstruction method, the pattern classification accuracy was calculated to be about 100 %. These techniques enabled the image acquisition of the central fuel rod and increased accuracy in partial defect detection. For further study, the performance of the proposed reconstruction method will be evaluated based on the experiment data.

This study explored soil samples in a depth about 10-15 cm, 20 Soil samples were collected systematically from various locations to ensure a comprehensive assessment of the site within Kasik oil refinery, Nineveh Governorate, Iraq, whereby the determination of three natural radionuclides' specific activities was carried out via A p-type coaxial HPGe γ-ray spectrometer, manufactured by CANBERRA, USA. Natural radioactivity in soil is a significant factor for assessing environmental radiation exposure and potential health risks. The ²²⁶Ra, ²³²Th, and ⁴⁰K radionuclides' mean specific activities in the soil samples were in the 13.10-33.43 Bq/kg, 10.58-22.76 Bq/kg, and 104.90-442.16 Bq/kg ranges, respectively, with overall means of 21.44 Bq/kg, 15.96 Bq/kg, and 255.50 Bq/kg, which are within the worldwide mean levels. The mean radium-equivalent activity (Ra_eq) was 63.94 Bq/kg, and thus below the 370 Bq/kg international limit. All of the samples' representative level index (I_γ) and internal and external hazard indices (H_in and H_ex) were below the limit of unity, indicating low radiation hazards. The absorbed dose rate resulting from the three primordial radionuclides was in the 19.92-41.46 nGy/h range (mean: 30.20 nGy/h), with an annual effective dose of 0.04 mSv/y. The mean excess lifetime risk of cancer was found to be 0.15 × 10^-3, and thus under the 0.29 × 10^-3 world average. The study's results indicate that the natural radionuclides' measured specific activities in the soil samples are below the world recommended values, suggesting that the investigated area is safe in terms of radiological health risks.

Preparations are currently underway to develop ISAA-Na code, a sodium-cooled fast reactors (SFRs) specific version of an integrated severe accident analysis code. This version aims to evaluate both protected and unprotected accidents in SFRs. Given the crucial role of fuel pin thermo-mechanical behavior in SFR safety analysis, an advanced model for MOX(Mixed OXide) fuel pin thermo-mechanical performance under operation conditions has been developed and validated. This validation involved simulating the SUPERFACT-1 fast reactor irradiation experiment using ISAA-Na. The resulting calculation data were compared with experimental measurements and the outcomes of other codes dedicated to fuel pin performance analysis (TRANSURANUS, MACROS, and GERMINAL) to assess the fuel pin thermo-mechanical model. The overall code assessment reveals that the model achieves satisfactory performance in simulating fuel pin behavior, with predicted errors falling within an acceptable range.

In this work, the source-terms such as nuclide inventories, radioactivities, neutron and gamma spectra, and decay heats were evaluated using new code BESNA in which the burnup dependent one-group cross sections were evaluated based on MCNP6. This work started with the selection of the reference spent fuels through a statistical analysis of all the accumulated PWR spent fuels in South Korea for safety analysis of a geological repository, which led to the two reference spent fuels for each of 16x16 and 17x17 fuel assembly types considering representative low and high burnups. Finally, the source terms were evaluated using our inhouse code BESNA for the reference spent fuels, and they were compared with those from ORIGEN in SCALE6.2 for validation purposes. The results of the comparison showed that BESNA gives good agreements with ORIGEN for all the considered source terms.

This paper describes the beam dynamic simulation with transfer matrix method for four sector cyclotron. Starting from a description on the equation of motion in the cyclotron, lattice functions were determined from transfer matrix method and the solutions for the 2nd-order nonlinear Hamiltonian were introduced and used in phase space particle tracking. Based on the description of beam dynamics in the cyclotron, simulation code was also developed for cyclotron design.

Scientific literature studies irradiated graphite treatment. Research also covers graphite conditioning and its long-term behavior under disposal conditions. The European Commission's CARBOWASTE project, titled "Treatment and disposal of irradiated graphite and other carbonaceous waste", is a key reference for state-of-the-art studies on alternative solutions. It identified 24 strategic options for managing irradiated graphite throughout its complete life cycle. The methodology proposed in this paper entails the application of the Analytic Hierarchy Process (AHP) method to rank the 24 options, placing particular emphasis on the weighting of seven criteria for selecting management options for the irradiated graphite. The highest weights were assigned by experts to 'environment and public safety' (28.05 %) and 'worker safety' (26.16 %). The objective is to develop a standardized approach enabling waste management companies to identify the most appropriate management option, considering structural and legislative constraints in their operating country. Examining the study findings, option 19 "In-situ entombment" stands out as the best choice in both the CARBOWASTE project and the proposed methodology. Thus, this methodology could assist hypothetical entities in examining management options for irradiated graphite, with the aim of identifying the optimal solution for graphite waste disposal.

The formation of scales in pipes is one element that has a major impact on the efficiency of machinery used in the oil and gas sector. With the help of artificial intelligence, this new, non-invasive device was able to figure out the volume fraction of a two-phase flow by taking into account the thickness of the scale in the tested pipeline. The proposed design uses an isotope pair of barium-133 and cesium-137 as a dual-energy gamma generator. One detector records photons that are transmitted, and another detector records photons that are scattered. The signals from the detectors were simulated using the Monte Carlo N-Particle (MCNP) code, and then ten frequency and wavelet characteristics were extracted. To choose the best inputs from the collected features for computing the volume fraction, an ant colony optimization (ACO)-based method is applied. Six attributes, representing the optimal combination, were developed using this method. In order to forecast the volume percentage of two-phase flows independently of flow regime and scale thickness, we fed the characteristics introduced by ACO into a group method of data handling (GMDH) neural network. Volume fraction calculations had a maximum RMSE of 0.056, which is quite little compared to previous research. By using the ACO to choose the best characteristics, the current work has significantly increased its accuracy in identifying volume fractions.

The stabilization technology for damaged nuclear fuels requires physically-stable pellets, so powders that are used for the process must have high sinterability. Due to safety limitations in H₂ concentration, fuel materials are reduced with high gas flow rates or for long times in cyclic oxidation and reduction process. We attempted to shorten the reaction time by shaking the powder reactor vertically. This method reduced the reaction time for the reduction process by two-thirds compared to that required in the static condition. The shaking process had a negligible effect on the quality of the powder. Therefore, this method can be applied to the reduction processes that require large volumes and long reaction times.

Realistic simulation of nuclear fuel performance requires not only validated models capable of describing the thermomechanical phenomena that take place within the fuel under irradiation conditions, but a detailed description of the thermal hydraulics of the channel surrounding the fuel rods, which provides the boundary conditions of the system. In this work, the main results and outlooks of coupling the thermal hydraulics code SubChanFlow with the fuel performance code DIONISIO are presented. To achieve this, an internal coupling was implemented, wherein DIONISIO is used as a master code controlling SubChanFlow as a thermal hydraulics subroutine replacing the simplified version already embedded in DIONISIO. Several tests were conducted to ensure the performance and quality of the coupling under normal operation conditions as a first approach. In addition, it was observed that the coupling demonstrated a significant improvement in the description of the cladding temperature and related variables, such as oxide thickness and hydrogen uptake, when compared with experimental data.

Matrix graphite is used as a structural material, thermal conductor, moderator, and secondary fission product barrier for fuel elements in high-temperature gas-cooled reactors (HTRs). Due to its high graphitization degree and compressibility, natural flake graphite (NFG) is used as the main filler in traditional A3-3 matrix graphite, whereas artificial graphite (AG), with a lower graphitization degree than NFG, serves as an additive for toughness and gas permeability. Matrix graphite could be improved in terms of thermal conductivity, oxidation resistance, and irradiation performance by increasing the degree of graphitization. However, reports on the development of new matrix graphite formulations are scarce. In this study, MG-20 matrix graphite was prepared by mixing 60 wt % NFG, 20 wt% natural microcrystalline graphite (MG), and 20 wt% phenolic resin. Due to the high graphitization degree (higher than AG) and low coefficient of thermal expansion (CTE) of MG, MG-20 exhibited higher thermal conductivity (~6%) and lower CTE (~2.4%) than A3-3. Thus, MG-20 with higher graphitization degree and better thermal properties than A3-3 could improve the performance of HTR fuel elements in the future.

A set of system performance tests has been performed and the thermal-hydraulic behaviors were investigated for the passive safety injection system (PSIS) of the SMART design. Major thermal-hydraulic characteristics of core cooling in reactor coolant system and safety injection (SI) in PSIS were investigated using the SMART-ITL. The parametric effects of break size, break location and train number on the PSIS performance were discussed with the relevant SMART-ITL data. Compared with the 2.0 inch SI line break test, the 0.4 inch SI line break test has a milder change of system parameters. The pressure is reduced more quickly, the decrease of water level is slower and the break mass flow rate is higher during the pressurizer safety valve line break than during the SI line break. Major thermal-hydraulic parameters were compared among single, two and full train operation of the PSIS focusing on the effect of train number. It was shown that multiple trains of CMT and SIT were operated independently and the reactor vessel (RV) inventory increased proportionally with the addition of each train. Sufficient SI water was injected into the RV in a passive manner to cool down the reactor core efficiently during the full train test.

Identifying important phenomena and parameters of LBLOCAs is an important step in nuclear power safety evaluation. Traditional identification processes are based on experience and incorporate important phenomena and parameters without quantifying specific aspects of the models. To accurately identify physical effects, this paper presents the use of multiphase field subchannel code for simulation analysis as applied to numerical examples and specific reactor thermal-hydraulic problems. The results indicate that the code can simulate the relevant phenomena, and the calculation band of the refill stage is narrow, while the calculation band of the reflood stage is wide. In addition, the key impact models are captured. The radiation model most significantly impacts the cladding temperature in the core, and the heat transfer model of transition boiling to the gas phase is most strongly influenced by the quenching time. The interfacial heat transfer model of the large bubble regime for the liquid plays a major role in influencing the water inventory of the lower plenum in the downcomer.

The approach for determining the break size of recirculation loops in a multiple-loop power plant in the event of a loss of coolant accident (LOCA) is presented in this study. In this study, the MAAP5 simulation program was used. An approach to measuring the size of a crack or break in the cooling system is the temperature difference between the recirculation loops. This method does not require any additional facilities; it compares the temperatures of the cooling loops to determine which one has a rupture. The best data source was the loop monitoring system, which sends temperature data for analysis to the main control room. A real operating power reactor training simulator and the FSAR are applied to evaluate MAAP5, the methodology's engine. The results of the MAAP5 simulation code were consistent with those of the power plant simulator. Therefore, MAAP5 could produce enough analytical data to create the relationship diagram between temperature difference and break size. The study hypothesized that there exists a maximum value of temperature difference corresponding to each break size and suggested that applying the absolute maximum temperature difference can aid in identifying the break size. This approach proposes an assistive method for determining the size of a fracture or break in the recirculation system by leveraging the temperature difference between each loop. This approach eliminates the need for additional facilities, as temperature data from the recirculation loops can be transmitted to the main control room. After the reactor scram, operators can monitor the maximum temperature differences at the inlet to estimate the break size. Although the fitting curve used to preliminary estimate the Large Break Loss of Coolant Accident break size may overestimate the break size, it still provides valuable insights. This novel tool offers a rapid and comprehensive method for detecting LOCA events in the recirculation loops.

By using the standard melt quenching method, samples of lithium borate glass-ceramics with a composition of 40Li₂B₄O₇ + xLi₂O +(50-x)NiO+5Gd₂O₃ + 5BaTiO₃ (x = 10 (C1), 20 (C2), 30 (C3), 40 (C4) and 50 (C5) wt%) have been manufactured. The structural and physical characteristics of the prepared C1-C5 samples have been investigated. Using an Ultra Ge detector and a 133Ba (3 Ci) radioisotope source as well as theoretically using EpiXS software, the radiation shielding competence of the C1-C5 samples were examined. The XRD analysis showed large crystal particle sizes were formed in the C1 sample, where the amount of NiO in its structure was maximum and the amount of Li₂O was minimum. The change in the NiO/Li₂O ratio caused the crystal particle size in the structure to decrease. Additionally, it was observed that the particle size increased in the C4 structure. In addition, the C3 and C5 structures became completely amorphous. This doping in the structure caused the crystal structure of C3 and C5 to deteriorate. While increasing the amount of NiO in the structure gave positive results up to a certain value, increasing the Li2O structure disrupted the crystal structure. A range of 2.91-4.51 g/cm³ was observed in the density of the synthesized C1, C2, C3, C4, and C5 samples. The dense C1 sample proved to have the largest mass-attenuation coefficient (MAC). The results obtained from experiments agree with the theoretical ones. There are no differences in the trends of the MAC and linear attenuation (LAC) coefficient. For each given energy, the half-value layer (HVL) organized as (HVL)_C5 > (HVL)_C4 > (HVL)_C3 > (HVL)_C2 > (HVL)_C1. Results revealed that glass-ceramics are one source of materials that can be used for radiation shielding.

The helical-coiled once-through steam generator (H-OTSG) has the advantages of compact structure and strong heat transfer ability, and is appropriate for lead-cooled fast reactor. The two-phase flow instability may cause mechanical vibration and thermal fatigue of heat transfer tube bundles, posing a serious threat to the safe operation of the H-OTSG. In this work, the secondary side of the H-OTSG characterized as multi-parallel channels is modeled by RELAP5/MOD3.4 code, and the oscillation behavior during start-up and the influence of structural and operating parameters on system stability are studied based on time-domain method. The results indicate that the pressure, flow rate, and temperature of the secondary fluid exhibit density wave oscillations at the heating section in a (n-2,2) pattern. In addition, increasing inlet throttling, reducing outlet throttling, controlling subcooling within an appropriate range and avoiding operation under low loads are all beneficial for improving the system stability of the helical-coiled once-through steam generator.

This study focuses on the Structure, System, and Components (SSC) of G.A Siwabessy Multipurpose Reactor (RSG-GAS), emphasizing the integration of reliability improvement within asset management optimization. The research aims to derive Maintenance Priority Index (MPI) values, contributing to system reliability ratings, crucial for assessing component functionality and estimating Remaining Useful Life (RUL) The methodology involves measuring critical aspects of system quality, safety, and cost. The MPI value, representing 10 % of SSC critical components, guides subsequent reliability calculations. Utilizing failure data from RSG-GAS components (2010-2018) and reactor core configurations (70th-96th core), SSC reliability is simulated using Monte Carlo simulation, with RUL based on real data. High MPI values are identified for critical components such as KBE01/AP01-02_B in the primary purification system, JE-01/AP01-02_A in the primary cooling system, and PA01-02/CR001_A in the secondary cooling system. Maintenance intervals of 100 days are recommended for KBE01/AP 01-02_B and JE-01/AP01-02_A, exceeding 50 % reliability, while PA01-02/CR001_A maintenance can extend to 75 days. Monte Carlo simulation results, with 1000 samples, closely align with real reliability values. The RUL result projects operational lifetimes of 225.6, 221.4, and 171.5 days for KBE01/AP 01-02_B, JE-01/AP01-02_A, and PA01-02/CR001_A components. This study improves asset management, offering practical insights for critical safety system component maintenance planning.

Medical radioisotopes (RIs) are widely used in the diagnosis and treatment of tumors. Theranostic RIs, such as ¹³¹I, ¹⁶⁶Ho, ¹⁷⁷Lu, and ¹⁸⁶Re, are particularly notable for their ability to enable diagnosis and treatment simultaneously. Conducting irradiation tests using the High-flux Advanced Neutron Application Reactor (HANARO), a 30-MW multipurpose research reactor, enables the production of these medical RIs. Prior to irradiation tests, it is crucial to assess whether they affect the thermal stability of irradiated materials and the thermal-hydraulic safety of HANARO. This study systematically investigates the feasibility of ¹⁶⁶Ho production using the isotope production irradiation hole of HANARO, focusing on thermal stability. The nuclear heating rates of the RI production target and RI capsule are calculated using MCNP6, and the calculated nuclear heating rates are used for three-dimensional heat transfer analysis using COMSOL Multiphysics. Under the assumed conditions in this study, ¹⁶⁶Ho production did not compromise the thermal stability of the RI production target and RI capsule; consequently, the thermal-hydraulic safety criteria of HANARO could be satisfied. This study can serve as a valuable reference for evaluating the thermal stability of irradiated materials and the thermal-hydraulic safety of HANARO, which must be performed before irradiation tests using HANARO.

The present study employed statistical methods to evaluate the possible radiological hazards linked to granitic rocks-bearing mineralization in the ELgarra region of Egypt. The geological structures influence the occurrence of uranium mineralization in this area and are primarily associated with altered granites. Gamma-ray spectrometry was utilized to examine the quantities of ²³⁸U, ²³²Th, and ⁴⁰K in granitic rock samples. The recorded levels of radioisotope activity concentrations in the analyzed regions ranged from 374 to 1740 Bq.kg^{-1 238}U, with an average of 1018 Bq.kg^-1. For ²³²Th, the range was between 71 and 163 Bq.kg^-1, with an average of 119 Bq.kg^-1. Lastly, for ⁴⁰K, the range was 756-1789 Bq.kg^-1, with an average of 1212 Bq.kg^-1. The detected levels of ²³⁸U, ²³²Th, and ⁴⁰K in the examined rock samples were observed to exceed the permissible limits of 35, 45, and 412 Bq.kg^-1, respectively. The primary radiological risks linked to these granitic rocks were attributed to the gamma rays released by the radioactive elements. Estimations of the radiological hazards in the granitic rocks were made, and statistical approaches were utilized to demonstrate the associations among radionuclides and radiological factors. The assessment confirmed that uranium, potassium, and their respective minerals in the granitic rocks were the key factors contributing to the radiological risks. As a result, the study determined that the granite rocks found in the study area needed precautions to be taken due to their high levels of radioactivity.

The current work investigates a computational model to study the thermal and hydraulic air behavior during the natural circulation at air ingression and accidents. This is done with the RHYS coupling ASYST VER 4 package. The test facility considered for the present study is a dual vertical channel facility comprised of two parallel channels connected to the upper and lower plenum. The flow fields in the heated and cooled channels were comprehensively characterized by analyzing axial temperature and velocity distributions using varied uniform iso-flux (100-1400 W/m²) and different outer surface temperatures (278, 288, 298, and 308 K). Temperature and velocity reversal recorded after maximal spots due to natural convection. The temperature rise from 278 to 308 K gave an average of 25.51 and 25.19° increase in air and inner wall temperatures, respectively, while air velocity increases at high cooling intensity (278 K) within the heated channel, in the cooled channel, low cooling intensity (308 K) resulted in higher velocity. The convective heat transfer is represented in terms of heat transfer coefficients, which are used to compute the Nusselt number. Additionally, the ASYST model was validated with data from literature sources, indicating strong agreement.

The present study focuses on investigating the gamma-ray protection features of clay bricks for potential use in radiation shielding fields. The study examined the physical and structural features that affect the performance of these stones in shielding γ-rays. The density (ρ, g/cm³) of the clay bricks samples was measured utilizing the MH-300A density meter. Additionally, the mineral structure within the annealed pressed clay samples was identification the XRD spectrometry. Moreover, the morphology and elemental chemical composition for the annealed bricks were examined using a Thermo Scientific Prisma E, USA field emission Scanning Electron Microscope (SEM) in conjunction with Energy Dispersive X-ray Spectroscopy. Besides, the shielding features of the clay bricks were analyzed using the experimentally measurements (by NaI (Tl) scintillation detector), XCOM software, and Monte Carlo Simulation over the γ-ray energy interval of 0.033-1.332 MeV. The findings of the study indicate that an increase in the pressure rate within the clay bricks samples leads to the rise in their density (from 1.62 to 1.87 g/cm³). This increase in density is accompanied by a decline in both porosity (Φ, %) (from 34.75 to 26.21 %) and water absorption (K, %) (from 26.21 to 14.74 %) factors. Furthermore, the increase in pressure rate from 7.61 to 114.22 MPa also results in an increase in the linear attenuation coefficient (μ, cm^-1) of the clay bricks under study. This is achieved by increasing the μ values from 0.39 to 0.43 cm^-1, from 0.13 to 0.15 cm^-1, and from 0.09 to 0.10 cm^-1, at 0.081, 0.511 and 1.173 MeV, respectively. The synthetic bricks offer a lead-free and efficient option for protection, making them ideal for use in nuclear facility start-ups or in areas with radiation exposure.