In order to achieve sustainable development that balances economic growth, environmental protection, and social well-being and ensures a sustainable future, strict environmental regulations and sustainable nuclear energy production may play a vital role. Empirical works are insufficient when measuring the effects of strict environmental policies and nuclear energy production on sustainable development. This research aims to close this gap by examining how environmental policy stringency and nuclear energy production contribute to sustainable development in the top 17 nuclear energy-generating countries between 1995 and 2021. The research uses the linear and nonlinear CS-ARDL and PMG-ARDL models to achieve this goal. The linear model suggests that environmental policy stringency and nuclear energy production contribute to long-term sustainable development. In the nonlinear model, a positive change in environmental policy stringency and nuclear energy production causes long-run sustainable development to grow, while a negative change in environmental policy stringency and nuclear energy production hinders long-run sustainable development. Furthermore, environmental technologies, human capital, financial development, trade liberalization, and research and development expenditures are crucial for fostering long-run sustainable development. In contrast, the natural resource rents hurt sustainable development. These findings suggest that policymakers should consider combining strict environmental regulations and nuclear energy in devising policies for sustainable development.

Compton imaging excels at visualizing gamma rays in the range of several hundred kiloelectronvolts to several megaelectronvolts. However, this technique has limitations in the imaging of low-energy gamma rays. In contrast, collimatorless imaging technique determines the location of a source by analyzing the distribution of interactions. Because the collimatorless imaging technique excels at imaging low-energy gamma rays that are easily shielded by detector components, it can compensate for the shortcomings of the Compton imaging technique. In this study, we propose a dual-mode imaging technique that selects the imaging method depending on the target gamma-ray energy and fuses them during reconstruction. The collimatorless imaging method demonstrated high angular resolution at low energy levels, whereas the Compton image surpasses it starting from 200 keV within its reconstructible range. The angular resolution of the dual-mode image was between those of the two methods. The trend of the positional error of gamma ray energy was similar to that of the angular resolution, and the dual-mode method exhibited the lowest average error of 0.7°. The dual imaging method exhibited higher efficiency, figure of merit, and signal-to-noise ratio by utilizing events from both imaging modalities. In addition, we investigated the geometrical effects of various structures.

In the city centre of Isparta, located in the southwestern part of Türkiye, indoor radon concentrations and gamma dose rate levels were observed above the world and country averages. For this reason, the determination of the natural radioactivity levels in soil samples of Isparta province is of great interest. In the present study, the activity concentrations of radionuclides (namely ²²⁶Ra,²³²Th, ⁴⁰K and ¹³⁷Cs) were investigated in the soil samples collected from 72 different points including the city centre and 12 towns of Isparta by using a gamma spectrometer with a high-purity germanium (HPGe) detector. The activity concentrations of ²²⁶Ra,²³²Th, ⁴⁰K and ¹³⁷Cs radionuclides in the samples were found in the range of 4.08 ± 0.39-69.84 ± 1.43 Bqkg^-1, 2.63 ± 0.28-65.66 ± 1.41 Bqkg^-1, 44.48 ± 3.17-452.38 ± 9.65 Bqkg^-1 and 0.28 ± 0.18-19.10 ± 0.61 Bqkg^-1 respectively. The averages of all the measurements were determined as 15.69 ± 0.72 Bqkg^-1 for 226Ra, 15.40 ± 0.78 Bqkg^-1 for ²³²Th, 210.60 ± 6.22 Bqkg^-1 for ⁴⁰K and 3.44 ± 0.31 Bqkg^-1 for ¹³⁷Cs. Although these values were below the country and world averages, it was revealed that the ²²⁶Ra and ²³²Th activity concentrations in the city centre exceeded the national and world averages. In addition, radiological risk maps of the analysed radionuclides were produced for the first time, covering all towns within the Isparta province. To evaluate the radiological hazards resulting from natural radionuclides, external hazard index (Hex), absorbed dose rate (D), annual effective dose equivalent (AEDE), male and female lifetime cancer risk (LCR) values were also calculated. The outcomes of this study revealed that all the radiological parameter averages for Isparta were lower than the world and country averages but the calculation results for the city centre displayed the opposite of this situation except for the external hazard index.

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation, radiative heat transfer, thermal ablation, and crust formation. To address the external vessel cooling, it is coupled with a conventional 1-D nuclear system analysis method. The 1-D system analysis code can calculate the two-phase natural circulation of cooling water and the convective heat transfer on the external reactor vessel wall. These two simulation codes exchange the temperature and heat flux of the reactor vessel outer wall. This study numerically simulated the IVR-ERVC strategy for a Korean high-power reactor and compared it with the traditional lumped parameter method (LPM). Unlike LPM, this study provides localized detailed data about the thermal hydraulic behavior of molten corium and visualization of phenomena in the IVR-ERVC strategy. This enhances our understanding of the phenomena in IVR-ERVC strategy and introduces new perspectives.

The PSA (Probabilistic Safety Assessment) is presented as a method of comprehensively evaluating the types of accidents that can occur at nuclear power plants. One of the key components to obtain technical success was the introduction of event tree analysis, and its strength and weakness has been reviewed and enhanced with the term, for instance, Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) since 1980s. However, the technical and policy maturity of IDPSA appears to have room for improvement: the terminologies need to be arranged internationally, and there are no official standards or guidelines for the method itself. Due to a misperception of IDPSA, there are even concerns that appropriate contribution in risk assessment cannot be made. For this reason, surveys and focus group interviews, sharing development experiences, and the direction of regulation and R&D on IDPSA were conducted targeting PSA experts in Korea. In order to support such process, authors have structured an overview of the development history and technical features of IDPSA. Finally, we will explore the ways to achieve synergy between the deterministic safety analysis and PSA, which may the origin of motivation how to deal with dynamic variability more properly such that an undue risk can be minimized.

A new automatic group structure optimization method based on the perturbation theory was proposed for the few-group structure in two-step nuclear design procedure for VHTR. It applies the sensitivity coefficient of the perturbation theory which includes not only the effect of the cross section on the multiplication factor but also the adjoint weighted reaction rate. The sensitivity coefficient of the fine group for the multiplication factor was calculated and the group boundary for a few-group can be determined so that the summation of the fine group sensitivity for a few-group should be evenly distributed over every few-group. This method was successfully implemented in the ABGO code. VHTR-350 and MiHTR 2D core were used to investigate the performance and applicability of the proposed method. The code generated the new group structures for two cores and the error of the multiplication and reaction rate by the new group structure was compared with the result by the fine group structure. The comparisons indicate that the new group structure by the proposed method can provide the multiplication factor and reaction rates comparable to the existing group structure and more accurate results than the group structure obtained using the Contributon theory.

This research intends to utilize melt-quenching technique in order to examine the radiation shielding capability of 10 % Mo, 10 % Na2O, 20 % PbO, and 60 % B₂O₃ glass system, with varying CaO, TiO₂, CuO, Fe₂O₃ or Mo. XCOM and MCNP simulations were utilized to analyze the radiation shielding properties of the fabricated glasses. The results revealed CuO having the superior MAC of 49.91 cm²/g, then Fe₂O₃ with 49.24 cm²/g, followed by CaO with 49.10 cm²/g, and TiO₂ with 48.49 cm²/g as the least. CuO and Fe₂O₃ were confirmed to have least HVL compared to CaO and TiO₂. The value of the lead equivalent thickness showed fluctuation against the gamma energy, where it raisess within the photoelectric region and falls after the photoelectric region. The data reveal that, the lead equivalent thickness at 0.1 MeV were 7.88 cm, 7.86 cm, 7.81 cm and 7.80 cm for TiO₂, Fe₂O₃, CaO, and CuO in the same order, respectively. The transmission factor (TF) raises as the gamma energy raises, having TiO₂ as the highest with 76.068 %, while the radiation protection efficiency dropped as the energy raises.

The selection of appropriate transport parameters of radionuclides is of utmost importance in the safety assessment of radioactive waste disposal sites. Both batch and column methods are widely performed to determine the transport parameters of radionuclides through soil layers. In this paper, the distribution coefficients K_d of Co ions in sandy mixtures containing soil samples taken from Anarak near-surface repository (Iran) were determined by batch and 3 different column methods namely, breakthrough curve, thin-section, and gamma spectroscopy methods. The distribution coefficients K_d of the mixture was obtained as 160 L kg^-1 for the batch method, 23 L kg^-1, 19 L kg^-1, and 18 L kg^-1 for the breakthrough curve, the thin-section, and the gamma spectroscopy methods, respectively. These results emphasize proper selection of transport data for using in the safety assessment programs.

In this article, we propose and explore a novel step in the digitization of the mapping of the spent fuel pool of nuclear power plants, in which the audio signal from the operator's microphone is used to obtain the identification codes of those components that are in each of the cells of the pool. In this way, we have not only an acquisition system but also a verification system that can be used in combination with the outcome of the analysis of the video signal. We developed an algorithm that uses at its core one of the latest models of multilingual Automatic Speech Recognition to transcribe audio signal, and with a post-processing of the timed transcriptions we build the identification code of fuel heads and other components. Results show a very high accuracy in audios from real recording of Spanish nuclear facilities, and the methodology proposed is easily extensible to other nuclear facilities in the world.

In the event of severe accidents in pressurized water reactor (PWR) nuclear power plants, the potential leakage of radioactive aerosols through containment cracks poses a considerable radioactive hazard to the public. Understanding aerosol transport and retention in cracks helps reduce the conservatism and uncertainty of radioactive hazard assessment. Concrete cracks are recognized as a pivotal pathway for the leakage of radioactive aerosols, and several researchers have undertaken experimental investigations concerning the aerosol transport and retention in concrete cracks. However, the majority of these studies have rather low gas flow Reynolds numbers. In this work, an experimental setup is built to study aerosol transport and retention in concrete cracks under high Reynolds number flow. The TiO₂ aerosol with a mass median diameter of 1 ㎛ and two concrete crack specimens are used in experiments. The results of gas flow experiments indicate that the Reynolds number is capable of reaching 10547. Combining the flow experimental data and Suzuki's formula, the equivalent heights of these two crack specimens are approximated as 303.67 ㎛ and 231.48 ㎛. The experimental results indicate a notably high retention rate of aerosols, exceeding 0.8. Furthermore, under high Reynolds number flow, the retention rate varies over a relatively narrow range, with the larger the equivalent height of the crack resulting in a lower retention rate. The experimental results match well with the mechanistic analysis based on inertial deposition theory, demonstrating the rationality of the inertial deposition theory.

The natural circulation system has been widely studied for use in various applications because of its inherent advantage. However, it has a key weakness called flow instability that makes the system unstable. Through massive previous research, the mechanisms of flow instability were analyzed, but there was an ambiguous aspect related to the effect of experimental parameters on the phenomenon. Particularly, there has been no report on the heat transfer performance of the system when flow instability phenomena were present. In this study, thermal-hydraulic phenomena of a two-phase natural circulation system that functions as a passive containment cooling system (PCCS) was investigated according to experimental parameters, namely, the temperature boundary (120-158 ℃) and exit loss coefficient (0-34.5) under atmospheric pressure conditions. The experimental results showed five different flow types in the loop. The flow modes that occurred by the interaction between flashing and boiling were classified by referring to the mass flow rate, void fraction, and visualization data. The system was more unstable when the temperature boundary conditions increased, but it was more stable when the exit loss coefficient increased. These results have only been confirmed in our research. The reason for the results is that the flow conditions are located on the boundary between Density Wave Oscillation I and the stable flow region, and that boundary does not have clear criteria. In addition, comparing the heat transfer performance of a system by heat rate can confirm the effect of flow instability on the thermal performance of the passive cooling system. As a result, the high exit loss coefficient stabilizes the system better than the low case and has similar heat removal performance.

Despite being developed more than four decades ago based on expert judgment, the THERP time-reliability correlation (TRC) remains widely employed for calculating diagnosis human error probabilities in human reliability analysis for nuclear power plant risk assessment. However, with numerous advancements in nuclear plant equipment and operations, as well as the emergence of plants featuring advanced interfaces, there's a growing need to validate the THERP TRC. The objective of this study is to establish a TRC for the diagnosis human error probability in a modern reference nuclear power plant equipped with up-to-date human-machine interfaces and compare it with the median of the THERP TRC. To achieve this goal, we devised a method to gather event diagnosis times from a simulator and developed procedures to derive diagnosis TRCs using this data. Our findings indicate that while the median of the THERP TRC offers a conservative diagnosis human error probability for up to 25 min, it becomes overly optimistic beyond this threshold.

A dosimetry system's ability to monitor absorbed doses is crucial for the quality assurance and control of products that have been exposed to radiation. Here, we propose a polyvinyl chloride (PVC)-methyl yellow (MY) film as a potential dosimeter for irradiation facilities. The PVC-MY film was prepared with a simple solvent-casting method at different MY concentrations from 0.3 to 1.2 mM. To evaluate the dosimeter's performance, the film was exposed to gamma-irradiation with irradiation doses of 0-25 kGy. The optical characteristics of the films examined using a UV-Vis spectrophotometer showed that the color of the films changed from yellow to red with increasing irradiation doses, and the color change sensitivity of the films increased with increasing MY concentrations. The absorption spectra of the irradiated PVC-MY films showed three maximum absorption peaks at 417, 522, and 547 nm. As irradiation doses increased, the film absorption at 417 nm decreased, while the film absorption at 522 and 547 nm increased. The film's stability studies suggest storage of the irradiated film in a cold-dark conditions is recommended to prevent damage and maintain response stability. The findings in this study emphasize the potential of the PVC-MY film as a dosimeter for irradiation facilities.

This research aimed to develop lead-free polymer composites based on poly-methyl methacrylate (PMMA) and embedded with tungsten carbide (WC) micro and nanoparticles for use in radiation protection applications. PMMA was filled with 20 %, 40 %, and 60 % by weight of WC micro- and nanoparticles. The shielding features of the proposed polymer mixtures were evaluated at different radioactive sources of different energies using the HPGe detector. The results revealed that the investigated composites containing micro and nano-structured WC particles showed superior radiation shielding at 81 keV due to the K-edge of the W element occurring at 69.5 keV. The findings also demonstrated that composites loaded with WC nanoparticles were more effective in shielding gamma radiation than those loaded with WC microparticles, even at the same filler wt.%. Furthermore, the sample containing 60 % by weight of nano-WC, coded as P-60nWC, had superior shielding performance than other polymer-based composites reported in the literature. Thus, the proposed nano-WC/PMMA composites can be effectively employed in radiation facilities as alternative environmentally and lead-free radiation shielding materials to protect people and the environment from the harmful risks of gamma radiation.

The swelling behavior of ferritic/martensitic steels (FC92-B/-N, HT9, and Gr.92) was investigated following 3.5 MeV Fe⁺⁺ ion irradiation. Tested alloys were helium-pre-implanted up to the peak contents of 120 and 240 appm with He/dpa ratio of 1 appm/dpa at room temperature and then exposed to self-ion beam to the peak damage conditions of 120 and 240 dpa at 475 ℃. Field-emission transmission electron microscopy was used to characterize the cavity evolution. FC92-B exhibited the highest resistance to swelling among the irradiated alloys. The final volumetric swelling of FC92-B reached 1.3 % at 70 dpa and 2.9 % at 140 dpa. On the other hand, HT9 exhibited the highest swelling, reaching 7.4 % at 140 dpa. Comparing the present swelling results at 140 dpa/140 appm He with swelling data at 280 dpa/280 appm He from our previous study, it was observed that Gr.92 and FC92-N swelled more at 140 dpa/140 appm He than at 280 dpa/280 appm He. This negative correlation between swelling and dose in Gr.92 and FC92-N is primarily attributed to the helium-associated swelling suppression at higher helium concentration of 280 appm. A bimodal cavity size distribution appeared only in Gr.92 and FC92-N at 280 dpa/280 appm. This result demonstrates that the excess amount of helium over 200 appm promoted early-stabilization of new-born cavities, resulting in preferentially enhanced cavity nucleation, while impeding the growth of nucleated cavities. An inhibition in cavity growth possibly led to an extended duration of nucleation-dominant stages, finally suppressing swelling in ion-irradiated Gr.92 and FC92-N alloys.

Noise reduction in low-dose positron emission tomography (PET) is a well-researched topic aimed at reducing patient radiation doses and improving diagnosis. Software-based noise reduction mainly improves the contrast between regions by reducing the variation of the acquired image. However, it should be performed under appropriate parameters to reduce discrimination. We propose a method that derives optimal noise-reduction parameters using the multi-scale structural similarity index measure and visual information fidelity, which are metrics for image quality assessment. Simulation and experimental studies demonstrated the viability of the proposed algorithm. The contrast-to-noise ratio value of the denoised reconstruction slice, which was used as the optimal parameter, increased approximately three times compared to that of the low-dose slice while preserving the resolution. The results indicate that the proposed method successfully predicted the parameters according to the noise-reduction algorithm and PET system conditions in the sinogram domain. The proposed algorithm should help prevent misdiagnosis and provide standardized medical images for clinical application by performing appropriate noise reduction.

Source term investigation is a critically important aspect of reactor decommissioning, particularly as the range of nuclides under consideration expands beyond the capabilities of existing analysis methods. In this study, we try to propose a methodology to indirectly determine the radioactivity of long-lived nuclides which are non-γ or low energy in various nuclear waste materials by measuring the radioactivity of ⁶⁰Co. The critical point of this method is to establish relationship between some easy to measure (ETM) key nuclides, such as certain γ emitters (like ⁶⁰Co), and the difficult to measure (DTM) nuclides to derive information for the DTM nuclides of interest. To begin, we calculate nuclide bulk densities of ⁵⁵Fe, ⁶⁰Co, ⁶³Ni and ¹⁵²Eu in nuclear waste materials. By constructing inversion matrices and analyzing the intensity matrices of characteristic γ lines emitted by ⁶⁰Co, we can extract the radioactivity of non-γ radionuclides (⁵⁵Fe, ⁶³Ni, and ¹⁵²Eu) present in the nuclear waste materials that are contained within a specific container. Furthermore, our methodology accounts for the influence of voids within the container, thereby ensuring the reliability and validity of the obtained results. This innovative approach offers a promising avenue for efficiently sorting nuclear waste.

The study focused on syenogranite and its altered variety, and specifically monitored the activities of ²²⁶Ra, ²³²Th, and ⁴⁰K using NaI(Tl) gamma-ray spectroscopy. Syenogranites in the study area in Egypt were investigated to determine their potential use in industries such as ornamental stones and ceramics. The main outcrops in the study area consist of metasediments, metavolcanics, metagabbro, syn-to late-orogenic granites, and post-orogenic granites (syenogranites). There are also numerous dikes and veins of various shapes and compositions that have intruded and penetrated all the older rocks in the area. The mean activity (± standard deviation, SD) of these radioelements in the granites is ²²⁶Ra (63 ± 71 Bq kg^-1), ²³²Th (76 ± 170 Bq kg^-1), and ⁴⁰K (1248 ± 532 Bq kg^-1) were found to be greater than the world average. The radiological hazards are mostly attributed to the γ-rays emitted by granitic rocks. The granites studied in this study deviate from international standards mainly due to the presence of minerals containing radioelements such as zircon, allanite, monazite, sphene titanite, apatite, thorite, samarskite, fergusonite, xenotime, columbite, apatite and fluorite. As a result, granitic rocks are unsuitable for home building.

Welding processes are used to connect several components in nuclear power plants. These welding processes can induce residual stress in welding joints, which has been identified as a significant factor in primary water stress corrosion cracking. Consequently, the assessment of welding residual stress plays a crucial role in determining the structural integrity of welded joints. In this study, a deep fuzzy neural networks (DFNN) with a rule-dropout method, which is an artificial intelligence (AI) method, was used to predict the residual stress of dissimilar metal welding. ABAQUS, a finite element analysis program, was used as the data collection tool to develop the AI model, and 6300 data instances were collected under 150 analysis conditions. A rule-dropout method and genetic algorithm were used to optimize the estimation performance of the DFNN model. DFNN with the rule-dropout model was compared to a deep neural network method, known as a general deep learning method, to evaluate the estimation performance of DFNN. In addition, a fuzzy neural network method and a cascaded support vector regression method conducted in previous studies were compared. Consequently, the estimation performance of the DFNN with the rule-dropout model was better than those of the comparison methods. The welding residual stress estimation results of this study are expected to contribute to the evaluation of the structural integrity of welded joints.

The Advanced Fuels Campaign's Fission Accelerated Steady State Testing (FAST) approach at Idaho National Laboratory creates a benchmark for evaluating accelerated irradiation via control rodlets and advanced metal fuel alloys for sodium-cooled fast reactors (SFRs). FAST experiments have been developed to generate prototypic temperature conditions during steady state irradiations of scaled geometric fuel pins. This approach helps to attain higher burn ups at a much faster rate than previous irradiation tests. For this study, the results from profilometry, fission gas release, and metallography of a FAST experiment are presented. Profilometry determined 0 % effective strain in the rodlets. The fission gas release fraction was measured from puncture/collection analysis. Constituent redistribution was observed in two specimens despite the peak fuel temperatures being below the normal ranges in which redistribution is expected. Metallography of the two higher temperature specimens showed typical swelling with the solid pin closing the fuel-cladding gap and the annular specimen having a fully closed annulus. Additionally, metallography indicated no swelling, no redistribution, and a homogenous microstructure for specimens with lower irradiation temperature. Post irradiation examination of FAST rodlets generally showed the expected representative behavior of metallic fuels within SFRs.

Traditional physical methods for measuring solubility were often affected by viscosity, density, and surface tension. In this study, we selected appropriate electrolytes to determine the solubility of impurities in the lead-bismuth circuit at reactor operating temperatures using electrochemical methods. Taking the measurement of the solubility of impurity Ce in liquid Bi as an example, we demonstrated that electrochemical methods can accurately determine its solubility in the LiCl-KCl molten salt. Furthermore, the successful measurement of other impurities such as La, Ce, U, and Mn in lead-bismuth eutectic (LBE) using the designed system has confirmed the universality of this method.

In this study, the groundwater hydrodynamics of two adjacent well sites were simulated under different pumping-injection ratios. The aim is to select an optimal pumping-injection ratio that can ensure the groundwater of the two well sites don't affect each other. In addition, the sulfur isotope composition of groundwater in the two well sites were analyzed to verify the simulated results. The results show that the flow velocity at different points outside the edge drilling hole decreases exponentially with the distance between the point and the edge hole. The streamline gradually extends outside of the borehole with the increase of leaching time. It is found that the optimal pumping-injection ratio is 1.003. In this case, the maximum distance between the moving front and the injection borehole is 28.44 m after leaching for 5 years. This indicates that the groundwater flow fields of the two well sites are well controlled. The significant difference in sulfur isotopes between the two well-sites further proves that the SO₄^2- in the acid mining zone does not affect the groundwater in the zone leached by CO₂+O₂.

New technologies in polymer synthesis and pipe extrusion equipment have led to the commercialization of high-performance, large-diameter, thick-wall high density polyethylene (HDPE) pipes. They have been used in the field of seawater transport and cooling to replace metal pipes, due to their advantages of high corrosion resistance and extensibility. Connection of HDPE pipe is important as it determines the safety of the entire piping system. Butt fusion welding is commonly used for HDPE pipe connection but may cause the formation of weak points in the welded joints, interfering the reliability of the pipeline system in the application of nuclear power plants. At present, there is a lack of research on evaluating the performance of welded joint for large-diameter thick-wall HDPE pipes made by butt fusion-welding. The purpose of this study is to investigate the influence of three different butt fusion-welding processes, i.e., single low pressure (SLP), single high pressure (SHP) and dual low pressure (DLP), by evaluating the performance of their welded joints, including characterizing tensile strength, extensibility, crystallinity and hardness. In specific, a thick-wall HDPE pipe with OD of 812.8 mm and wall thickness of 74 mm which is certified for nuclear safety class was used for study. Representative specimen from the outer, middle and inner part across the wall of the main pipe body and welded joints were taken for testing. Different test methods and specimens were designed to assess the feasibility of evaluating the welding performance from different welding process. The results showed that the mechanical properties of different locations of the welded joints were different, and the tensile strength and fracture energy of the middle part of the joint were lower than that of the inner and outer parts, which could be caused by the difference in the crystallinity and thickness of the melting zone influenced by welding processes, as can be seen from the analysis of DSC test and morphology observation. Hardness testing was conducted on the section of the welded joints, and it revealed that the micromechanical properties of the welded joints in the region of the heat-affected zone were enhanced significantly, which may be due to the annealing effect caused by welding process. In summary, The DLP process resulted in the best extensibility of the welded joints among three processes, suggesting that the joining pressure from welding process plays an important role in affecting the extensibility of the welded joints.

Compared to traditional transverse integration methods, the variational nodal method, with its unique advantages, is more suitable for high-fidelity calculations of reactor physics in reactors with complex geometries and finer detail descriptions. In this study, the basic theory of the variational nodal method was derived and the VINUS code is developed. The neutron solver based on this method is adaptable to various geometric models, and showcased the code's fundamental framework. On this basis, a set of self-designed macroscopic cross-section benchmarks, actual macroscopic cross-section benchmark VVER-440, and few-group microscopic cross-section benchmark RBEC-M for fast spectrum reactors were used to verify different functionalities of VINUS. The results were shown that VINUS code maintains good computational accuracy and convergence trends. For the VVER-440 benchmarks, the deviation of k_eff of VINUS from reference is less than 100 pcm, and the maximum power deviation is less than 4 %. For the RBEC-M, the deviation of k_eff is 125 pcm, and the maximum power deviation is less than 5 %. These outcomes collectively demonstrate the solver's potential for engineering applications in future advanced reactor designs.

One of the essential requirements for molten salt reactor (MSR) design is methodology for analyzing multiphysics phenomena, such as the behavior of liquid fuel. In the research of Molten Salt Fast Reactors (MSFRs), the Neutron Diffusion Equation (NDE) is widely employed. This study introduces a method to enhance the accuracy of neutronic analysis of MSFR using the NDE. Using the simple structure of MSFR and the characteristics of liquid nuclear fuel, it is intended to enable the application of the simple equivalence method, which is difficult to perform on the existing fast reactor. A straightforward yet effective approach named the simplified Generalized Equivalence Theory (simplified GET) is proposed for applying flux-volume-weighted homogenized cross-sections and representative Discontinuity Factor (DF) values obtained at material. This approach, while similar to the General Equivalence Theory (GET) method, significantly simplifies the enhancement of accuracy in reactor analysis, minimizing computational efforts. Our investigation spans from the initial core to the burned core, ensuring the applicability of this simple method throughout the reactor's operational life. The proposed method demonstrates promising results, offering a viable solution to improve the accuracy of NDE-based calculations in MSFRs.

We present a concise equation correlating the burnup of spent nuclear fuel (SF) with the neutron count rate, developed through comprehensive data analysis from Origen-ARP and MCNP codes. This equation is applicable to Fork detectors, commonly used for verifying SF assemblies before their transfer to new storage sites. The detector is assumed to be a helium-3 detector. The reaction rate (RR) in the ³He detector is influenced by Total Neutron Source Intensity (TNSI), net neutron multiplication, and neutron capture during transit to the detector. TNSI emerges as the most influential factor. Two scenarios were explored: one involving pure water and the other with water containing 2000 ppm of boron. The characteristics of the concise equation are also analyzed.

Nuclear energy is used more and more widely as a clean energy source, but nuclear energy facilities are risky, and when a nuclear leak occurs, it is necessary to detect equipment in a nuclear radiation environment. In the nuclear radiation environment, due to the impact of high-energy particles on the camera sensor, the collected image contains a lot of radiation noise, which greatly reduces the visual perception of the image. Aiming at the problem that radiation noise reduces image quality, a radiation image compound filtering algorithm combining median filtering and multi-frame average filtering is proposed based on radiation noise characteristics. Compared with several common filtering algorithms, the radiation noise image is filtered under the same radiation dose and achieve the highest Peak Signal Noise Rate (PSNR) and Structural Similarity (SSIM), and compared with the multi-frame average filtering method, the number of image frames required by the algorithm in this paper is greatly reduced. Experimental results show that the algorithm can effectively eliminate radiation noise and is more suitable for image filtering in radiation environment.

Using strong electromagnetic fields generated by lasers to interact with electrons for precise diagnosis and manipulation of electron beams represents a recent focal point in accelerator technology. This approach surpasses the limitations of conventional RF technology, such as low electric field gradients and timing jitters, effectively enhancing the accuracy of ultrafast electron beam diagnostics and manipulations. As demands for precision continue to rise, the precise diagnosis of crucial parameters of ultrafast electron beams remains challenging. This study delves into the electromagnetic behavior of THz-driven devices and proposes an all-optical method utilizing single-cycle THz radiation to compress and characterize a 3 MeV electron beam. Particle tracking simulations demonstrate an astonishing compression effect, reducing the bunch length from 54.0 fs to 4.3 fs, and achieving sub-femtosecond bunch length measurement resolution. Moreover, when combined with an orthogonal THz streak camera, this method shows even greater potential in multi-bunch scenarios.

We propose a new method of X-ray backscatter imaging which is a significant development of the well known flying spot approach. Instead of sweeping the imaged object by pencil X-ray beam, and recording the amplitude of backscattered radiation in order to form the raster image, the object is irradiated with a fan beam of X-rays, forming a line on the imaging plane. By properly chosen motion of the collimating slit, a set of lines, which makes up a 2D sinogram of the original object, can be collected. The backscatter image can then be obtained using one of the tomographic reconstruction algorithms, such as the filtered backprojection, which is well known in computed tomography or nuclear medicine. The feasibility of the method has been demonstrated experimentally using a prototype scanner with an industrial X-ray source. The main advantage of our method is the essentially more efficient use of the available X-ray source beam, by illuminating the object with a substantially larger photon flux at each time point. As a result, the image can be obtained at much shorter acquisition time and/or at much lower source power.

Reinforced concrete structures are subject to exposure to chloride ions in the air, leading to chloride penetration, and carbonation attacks resulting from exposure to carbon dioxide. This chemical degradation process induces corrosion of reinforcing bars within concrete, significantly impacting durability. Structures situated in coastal areas, such as nuclear power plants, are particularly susceptible to rapid chloride penetration due to the high chloride concentration in the air. This study utilizes existing experimental data to forecast the chloride diffusion coefficient employing artificial neural network (ANN technology). The total number of experimental data was 535 gathered from 18 papers. Through analysis of the chloride coefficient and predicted degradation depth, the effective cross-sectional area of concrete is examined, and the deterioration of wall performance is forecasted.

The small modular PWR (SMPWR) usually adopts integral design. Under severe accident, the system responses are different from those large PWRs. It is necessary to study the severe accident behavior of the SMPWR. A MELCOR model is developed for SMPWR and its steady-state results are in good agreement with the design values. Severe accidents induced by secondary pipeline break accidents are simulated, and no pressure relief measures are taken to keep the primary loop under high pressure. The mitigation effects of passive containment air cooling system (PAS) and passive cavity injection system (PCIS) are evaluated under different cases. The results show that under high pressure conditions, PCIS can effectively cool the lower head. The earlier the PCIS operates, the more significant the mitigation effect can be. In addition, PAS can effectively reduce the peak pressure and temperature in the containment. This study can provide a reference for the formulation of severe accident management guidelines on SMPWRs.

This research article introduces a study regarding the uncertainties treatment during severe accidents for Pressurized Heavy Water Reactors (PHWRs). The present study is focused upon the unmitigated Station BlackOut (SBO) accident analysis for a CANada Deuterium Uranium (CANDU) type reactor emphasizing the impact of the uncertainties treatment on the relevant key timings of the SBO accident progression through different approaches for the uncertainty parameters' Probabilistic Distribution Functions (PDFs). A comparison between the sensitivity analysis results is provided in the present research study. The uncertainty analysis is performed with the RELAP/SCDAPSIM code with the Integrated Uncertainty Analysis (IUA) package from the code. Results from the research would support the advancements on the best-practices for uncertainty analyses with respect to the parameter's uncertainties distribution functions. Data dispersion is a key element for the realistic quantification of uncertainties in nuclear power plant safety analyses, including severe accidents.

The binary decision diagram (BDD) method provides a means to calculate the exact probability of a fault tree, but cannot solve large fault trees for nuclear power plant PSAs. However, the BDD method can be used as a supplementary means to increase the accuracy of PSA quantification, especially for a seismic PSA which includes many non-rare events. Since it is difficult to solve a large PSA model using BDD method, several approaches have been developed that partially apply the BDD technique to small-sized branches of a large PSA model. This paper proposes an approach of partially applying BDD technique to large-sized branches. The results from the pilot application to a seismic PSA model shows that the partial BDD approach of this paper provides a relatively accurate way to quantify each sequence as well as the overall core damage in a seismic PSA model.

The Fukushima accident showed that the safety of multiple nuclear power plants (NPPs) at the same site could be jeopardized simultaneously. Since then, many studies have focused on developing strategies to prevent the spread of multi-unit accidents, with numerous countries establishing strategies to use mobile equipment. However, mobile equipment strategies are inherently accompanied by a high degree of uncertainty regarding operation success and duration because multiple organizations and personnel interact in various ways during multi-unit accident situations. Furthermore, supplementing current fixed equipment with additional mobile equipment requires extra resources. Therefore, cross-tie strategies that use currently installed fixed equipment can provide additional means to manage site risk with relatively few additional costs. This study proposes a multi-unit probabilistic safety assessment-based risk evaluation framework for utilizing cross-tie systems in NPPs and a modeling methodology to quantify the effectiveness of the cross-tie strategies. A case study was conducted to evaluate the risk reduction from using cross-tie strategies for emergency diesel generators and alternate AC diesel generators, which are power systems utilized in multi-unit loss of offsite power initiating events. It is expected that the developed framework and methodology can be utilized for other types of cross-tie strategies as well.

This study investigates Equivalent Spherical Diameter (ESD) estimation at high inlet velocity pool scrubbing conditions using the Interfacial Area Transport Equation (IATE) diameter model including bubble-induced turbulence and interphase modeling. The compatibility of area-averaged Sauter Mean Diameter (SMD), areaaveraged Local Equivalent Diameter (LED) and void-weighted area-averaged LED approaches to estimate the ESD are explored and the proposed model is validated against available experimental data. The study reveals that the prevalent constant ESD assumption in pool scrubbing codes is not universal by showcasing a decreasing trend along the column due to intensive bubble breakup. The area-averaged LED approach fails to capture this trend, while the area-averaged SMD and void-weighted area-averaged LED approaches provide accurate estimations aligned with experimental data. Turbulence parameters, interfacial forces, and diameter modeling are identified as crucial for accurate predictions of flow and geometrical variables by setting up the OpenFOAM framework. A sensitivity analysis indicates that the inlet velocity has an acceptable effect on the ESD along the column. The ESD increases near the exit and decreases in the swarm region by increasing the inlet velocities. Turbulent intensity reduces ESD across all column sections while changes in aspect ratio minimally impact ESD. The study shows promise in developing correlations that take into account the spatial variation of ESD in pool scrubbing conditions.

Preventive Maintenance(PM) for safety component during power operation at nuclear power plants, On-Line Maintenance(OLM) refers to intentionally entering the Limited Condition of Operation(LCO) specified in the Technical Specification(TS) for safety-related systems and components in order to perform preventive maintenance within the Allowed Outage Time (AOT). This study assessed the feasibility of conducting OLM at the domestic APR1400 nuclear power plant. It focused on preventive maintenance duration and risk perspectives. A total of 78 FEGs were developed for 4450 facilities, considering system functions and preventive maintenance scope during output operation for eight safety-related systems. Additionally, maintenance items included in FEGs were selected, designated as targets for OLM, and their maintenance durations were evaluated and compared with AOT for each maintenance item. As a result, the Auxiliary Feedwater and Essential Chilled Water systems were identified as systems allowing OLM. Furthermore, utilizing the Risk Monitoring System (RIMS), the increased risk value due to the unavailability of target equipment during preventive maintenance was analyzed to determine whether it falls within the acceptable range. Regarding the temporary risk increase caused by OLM, it was observed that in all systems, it falls within Zone III according to NUMARC93-01 standards, allowing for normal equipment arrangement for OLM. However, according to the risk increase standards rate in domestic nuclear power plants, when maintaining the A-train in four systems including Component Cooling Water, they are all evaluated as 'Orange,' indicating that measures for risk mitigation are necessary for OLM to be feasible. When considering extending AOT up to 1.6 times the maintenance time, the risk increase falls within Zone III according to permissible change in risk standards, indicating that AOT extension might be feasible based solely on risk changes. To apply OLM within the permissible risk management scope in domestic nuclear power plants, regulatory policies need to allow voluntary LCO entry for preventive maintenance, necessitating clear determination by regulatory agencies using risk-informed policies. While OLM seems viable concerning maintenance duration and quantitative risk aspects, for inducing regulatory policy changes, comprehensive OLM guidelines are necessary, including risk management strategies.

Some core designs integrate high-enriched fuel and moderator materials to enhance neutron utilization. This combination results in a broad spectrum within the system, posing challenges in resonance calculation. This paper introduces a general framework to realize resonance self-shielding treatment in broad-spectrum fuel lattice problems. The framework consists of three components. First, a new energy group structure is devised to support resonance calculation in the entire energy range and capture spectral transition and thermalization effects during eigenvalue calculation. Second, the subgroup method based on narrow approximation is selected as a universal method to perform resonance calculation. Finally, transport equations for each fissionable region are solved for neutron flux to collapse the fission spectrum. The proposed method is verified against fast, intermediate, and thermal spectrum pin cell problems and an assembly problem featuring a fast-thermal coupled spectrum. Numerical results affirm the accuracy of the proposed method in handling these scenarios, with eigenvalue errors below 154 pcm for pin cell problems and 106 pcm for the assembly problem. The verification results revealed that the proposed method enables accurate resonance self-shielding treatment for broad-spectrum problems.

Fluid pipelines, commonly utilized in the oil industry, often face efficiency and reliability issues due to sediment buildup causing erosion, corrosion, and pipe wall thinning. Traditional assessment methods involve disruptive measures like cutting or creating holes and temporarily taking pipelines out of service. A non-destructive alternative, Limited-Number-Detector Computed Tomography (LNDCT), proves cost-effective and superior. Our proposed algorithm enhances data acquisition and projections using discrete detectors, employing Co-60 as a gamma-ray source and thallium-doped sodium iodide, NaI(Tl), detectors in an arc configuration. Monte Carlo simulations aligned closely with experimental data. Optimization involved adjusting the detector aperture angle based on a primary-to-scatter ratio of gamma-ray photons. We investigated the utility of various isotopes (Co-60, Cs-137, Am-241, Ir-192) to determine optimal projection signal amplitude. The algorithm generates a large sinogram matrix, and a filtered back-projection algorithm with a Hamming filter maximizes image quality while ensuring acceptable calculation volume and time. Using four phantoms, including pipelines filled to different scales, our study evaluates LNDCT configuration, performance, and validation. The results highlight its potential for efficiently evaluating sediment in pipelines, confirming the correctness and accuracy of our proposed algorithm.

Utilizing first-order beam dynamics models is adequate for studying the beam properties during the conceptual design of a cyclotron-based proton therapy beamline. After finishing lattice design, particle-matter interaction simulations for passive elements (e.g., degrader, collimators, energy slit) are required. The cascade simulation is used for lattice updates in each iteration, which is complicated. In addition, when the models involve particle tracking and particle-matter interaction, their optimization process is time-consuming. Therefore, this study proposes a start-to-end modeling method using Monte Carlo Beam Delivery Simulation (BDSIM) software that considers more realistic factors, such as particle-matter interaction and the realistic vacuum chamber, to precisely evaluate working parameters, along with an efficient optimization method that utilizes multi-objective Bayesian optimization (MOBO) to improve transmission efficiency. Taking the Huazhong University of Science and Technology proton therapy facility (HUST-PTF) as an example, beam loss along the beamline is located, quantified, and subsequently reduced by tuning the quadrupole strengths based on MOBO. The results show that: (i) By considering the particle-matter interaction and the realistic vacuum chamber, the precision in the prediction of the beam properties is improved; (ii) After optimization, the transmission efficiency of the entire beamline is relatively increased by an average of 6.52 % under different energy settings, especially 11.39 % at 70 MeV.

In this study, we first produced the Korean-specific dose coefficients (DCs) for soil contamination using the Mesh-type Reference Korean Phantoms (MRKPs). The Korean DCs were compared with the values in ICRP Publication 144 produced using the Caucasian-based ICRP reference phantoms to investigate dosimetric impact due to the racial difference (Korean/Asian vs Caucasian). Monte Carlo dose calculations using the Geant4 code were conducted where the photon and electron sources in the phase-space data used for the ICRP-144 DC calculations were irradiated to the MRKPs. For photons, the organ DCs of the MRKPs showed a good agreement with the ICRP-144 DCs (deviations <20 %) for most energies, while significant differences at energies below 0.05 MeV were observed by up to a factor of 55.6 (thymus at 0.015 MeV). For electrons, notable differences in the organ DCs were observed the overall energy region (deviations >20 % for most cases). The effective DCs of the MRKPs showed an excellent agreement with the ICRP-144 DCs for photons (deviations <16 %), whereas notable differences by up to 1.7 times (0.05 MeV) were observed for electrons. The Korean DCs for soil contamination will be beneficially used in dose estimates for Koreans especially in risk assessments.

NORM is used as a raw material in various industries worldwide. According to the IAEA, radiological effects may occur in the environment due to raw materials and by-products generated from NORM industries. The objective of this study was to assess radiological environmental impact of the phosphate processing industry to identify the radiological effects on the general public. The resident farmer scenario was chosen as the exposure scenario due to the living characteristics of the public around the facility and the conservatism of the assessment. The RESRAD-OFFSITE code was used to evaluate the radiation dose to the public. The maximum radiation dose to the public was 6.13 × 10^-3 mSv/y. Main exposure pathways were aquatic food ingestion, radon inhalation, and meat ingestion. The uranium series accounted for about 99 % of the total radiation dose, while the thorium series and K-40 nuclides accounted for less than 1 %. These study results can be used for management of radiological impact to the public around domestic NORM industries.

The response of an anti-Compton phoswich detector to gamma rays was investigated using Monte-Carlo method, and the pulses from different crystal cases, including gamma deposition only in the LaBr3(Ce) or CsI(Tl) crystal and coincidence in both crystals, were analyzed. A novel pulse discrimination method for gamma deposition events based on wavelet transform analysis, called SSD (Scale Shape Discrimination), was developed in this study. Compared to the traditional PSD (Pulse Shape Discrimination) method, SSD has the advantage of transforming one-dimensional pulses in the time-domain into two-dimensional time-frequency spectra, providing the more useful features for pulse discrimination. The performances of the Compton suppression and Full-energy peak loss using PSD and SSD methods was studied. The results show that the Compton suppression factor I_PSD = 5.12 and I_SSD = 5.32, and FEP loss factor PL_PSD = 0.0554 and PL_SSD = 0.0587. Meanwhile, the influences of the cutoff values for pulse discrimination on the results of I and PL with different method were analyzed.

The multi-dimensional thermal-hydraulic phenomena in the downcomer of advanced pressurized water reactor with direct vessel injection system are the key points for the safety analysis during a loss of coolant accident. In order to improve the accuracy of LOCUST code for the predictions of thermal-hydraulic phenomena in downcomer region, some newly correlations have been implemented into LOCUST code. The wall friction model of LOCUST code was modified based on the correlations which developed by Yang. The interfacial friction models in LOCUST code have been modified as Hibiki-Ishii correlations. In addition, in order to simulate the upward flow of recirculation flow in downcomer region, the Kinoshita-Hibiki correlations have been also implemented into LOCUST code for better simulating the recirculation flow in downcomer region. The modified code was validated with experimental data of DOBO facility. Five tests of DOBO facility have been calculated by LOCUST, and the calculated axial void fraction distributions have been compared with the measurements. The results show that the modified LOCUST with new correlations of distribution parameter and drift velocity shows better accuracy than the original code. The deviations of the modified LOCUST code are less than the original code and are almost within ±20 %.

The dissolved oxygen in a coolant can affect the oxidation properties of structural materials. A desirable oxide phase formation is achieved by manipulating the oxygen level in the coolant, which can mitigate structural material degradation in nuclear power plants. Therefore, the role of dissolved oxygen in the corrosion of structural materials in aqueous environments needs to be understood. In this study, a short-term corrosion test (up to 300 h) of Ferritic/Martensitic steels (F/M steels; FeCrW model alloys), namely, Fe12Cr1W, Fe9Cr1W, and Fe9Cr, in stagnant water at 360 ℃ was performed in a pressurized autoclave with the dissolved oxygen concentration controlled to 1 ppm or a very low level (<1 ppm). The results of the corrosion tests showed that an increase in the oxygen level in the water elevated the corrosion potential, allowing the phase transition of iron oxide from magnetite (Fe₃O₄) to hematite (Fe₂O₃), whereas there was no significant correlation between the concentrations of the alloying elements Cr and W and the oxide growth rate. In addition, hematite was found to mitigate further oxide growth. Finally, a mechanism for the growth of the initial oxide layer was proposed based on the experimental results.

The CFETR multipurpose overload robot (CMOR) is a critical component of the fusion reactor remote handling system. To accurately calculate and visualize the structural deformation and stress characteristics of the CMOR motion process, this paper first establishes a CMOR kinematic model to analyze the unfolding and working process in the vacuum chamber. Then, the dynamic model of CMOR is established using the Lagrangian method, and the rigid-flexible coupling modeling of CMOR links and joints is achieved using the finite element method and the linear spring damping equivalent model. The co-simulation results of the CMOR rigid-flexible coupled model show that when the end load is 2000 kg, the extreme value of the end-effector position error is more than 0.12 m, and the maximum stress value is 1.85 × 10⁸ Pa. To utilize the stress-strain data of CMOR, this paper designs a CMOR morphology prediction control system based on Unity software. Implanting CMOR finite element analysis data into the Unity environment, researchers can monitor the stress strain generated by different motion trajectories of the CMOR robotic arm in the control system. It provides a platform for subsequent research on CMOR error compensation and extreme operation warnings.

The development and assessment of pharmaceutical dosage forms make considerable use of gamma-scintigraphy. Gamma scintigraphy is an imaging technique that is integrated with CT to assess and evaluate the targeting of drugs to various delivery sites, the impact of treatment, and the severity of the disease. A small number of radioisotopes were tagged with the delivery system and emitted radiation can be visualized by the gamma camera which forms a 2-D image displaying the tissue-specific distribution of radioactivity. The isotopes that are used widely include Technetium-99 m (99Tc), Iodine (131I), Fluorodeoxyglucose (18F-FDG), Fluoromisonidazole (18F-FMISO) and Gallium (Ga67), Indium (111In). This review mainly covers different applications of gamma scintigraphy for the assessment of drug targeting via different routes to different organs and their visualization by gamma scintigraphy. The review mainly focuses assessment of drug targeting in the tumor tissue, thyroid gland, brain, pulmonary pathway, skin deposition, detection of renal impairment as well as cardiac diseases, drug release from formulations, drug deposition in arthritis, drug retention in the scalp, and behavior of formulation when administered via intra-vaginal route. Various pre-clinical and clinical studies were included in the review that demonstrates the importance and future of gamma scintigraphy in sensing drug delivery.

During the process of bilateral teleoperation, operators not only need to perform complex maintenance tasks but also have to constantly monitor the safety of the operation, leading to reduced operational efficiency. Therefore, in this paper, we introduce a shared autonomous scheme that intervenes in the operator's command input when necessary, autonomously ensuring the safe operation of the manipulator by employing a rolling horizon planning controller based on Nonlinear Model Predictive Control (NMPC). This controller considers the motion boundaries and collision avoidance constraints of the manipulator, accompanied by the design of corresponding objective functions. To validate the effectiveness of the proposed method, we conduct tests on collision-free trajectory tracking and comprehensive performance with collision constraints, confirming the feasibility and excellent performance of the approach.

This paper presents a seismic analysis method that can evaluate a very large number of cases for the free standing dry storage cask by proposing a methodology that has short analysis time as well as accuracy. This study also performed a seismic analysis of a dry storage facility with multiple casks to show a tip-over phenomenon from earthquake accident conditions. The earthquake accident condition is long-term event that occur during about 20 s long, and lots of seismic analysis cases should be performed to consider various real conditions because the free-standing spent-fuel dry storage cask has many nonlinear responses. The soil-concrete pad-cask interaction was considered in the seismic analysis and finite element model was made using concrete pad, soil and cask models. In the reinforced concrete pad, the rebar was excluded to reduce the analysis time, but the thickness was corrected to maintain the bending rigidity. Additionally, the analysis time reduced by modeling the cask as a rigid body rather than a flexible body. 35-cases of seismic analysis were performed to determine a tip-over phenomenon from each earthquake. The analysis revealed that no tip-over phenomenon of the cask was observed in all analyses from 0.2 g to 0.6 g, however the tip-over of the cask were observed from 0.8 g with friction coefficients of 0.8 and 1.0.

This study investigates the efficient passivation of CdTe-based semiconductor crystals, focusing on the long-term stability and underlying mechanisms of NaOCl passivation. CdZnTeSe crystals were grown, processed, and passivated with NaOCl, and their surface characteristics were studied using SEM-EDS and XPS. The passivation was found to significantly enhance the surface resistance, with a sustained effect for more than 90 s, attributed to the formation of a tellurium oxide layer. The passivation process was further elucidated through detailed morphological and compositional analyses. The NaOCl-passivated crystals exhibited improved electrical and spectroscopic properties in radiation detection, with a prolonged stability of 60-90 days, which are longer compared to other passivants. Additionally, the feasibility of NaOCl passivation on a CdZnTeSe detector was explored, showcasing enhanced material resistance and spectroscopic performance. The study concludes with insights into the potential industrial application of NaOCl passivation for CdTe-based radiation detectors.

This paper details a computational framework to produce automated, graphical workflows, and how this framework can be deployed to support complex modeling problems like those in nuclear engineering. Key benefits of the framework include: automating previously manual workflows; intuitive construction and communication of workflows through a graphical interface; and automated file transfer and handling for workflows deployed across heterogeneous computing resources. This paper demonstrates the framework's application to probabilistic post-closure performance assessment of systems for deep geologic disposal of nuclear waste. However, the framework is a general capability that can help users running a variety of computational studies.