• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1310-1319
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• 2024

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models' execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system's size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier-Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.207-215
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• 2024

This study focuses on measuring the levels of naturally occurring radioactivity in the soil of Swabi, Khyber Pakhtunkhwa, Pakistan, as well as the associated health hazard. Thirty (30) soil samples were collected from various locations and analyzed for ²²⁶Ra, ²³²Th, and ⁴⁰K radioactivity levels using a High Purity Germanium detector (HPGe) gamma-ray spectrometer with a photo-peak efficiency of approximately 52.3%. The average values obtained for these radionuclides are 35.6 ± 5.7 Bqkg^-1, 47 ± 12.5 Bqkg^-1, and 877 ± 153 Bqkg^-1, respectively. The level of ²³²Th is slightly higher and ⁴⁰K is 2.2 times higher than the internationally recommended limit of 30 Bqkg^-1 and 400 Bqkg^-1, respectively. Various parameters were calculated based on the results obtained, including Radium Equivalent (Ra_eq), External Hazard (H_ex), Absorbed Dose Rate (D), Annual Gonadal Equivalent Dose (AGDE), Annual Effective Dose Rate, and Excess Lifetime Cancer Risk (ELCR), which are 170.3 ± 24 Bqkg^-1, 0.46 ± 0.06 Bqkg^-1, 81.4 ± 2.04 nGy h^-1, 582 ± 78.08 µSvy^-1, 99.8 ± 13.5 µSv Gy^-1, and 0.349 ± 0.04, respectively. These values are below the limits recommended by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) in 2002. This study highlights the potential radiation threats associated with natural radioactivity levels in the soil of Swabi and provides valuable information for public health and safety.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.301-308
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• 2024

Ceramic materials are being explored as alternatives to toxic lead sheets for radiation shielding due to their favorable properties like durability, thermal stability, and aesthetic appeal. However, crafting effective ceramics for radiation shielding entails complex processes, raising production costs. To investigate local viability, this study evaluated Malaysian ceramic tiles for shielding in diagnostic X-ray rooms. Different ceramics in terms of density and thickness were selected from local manufacturers. Energy Dispersive X-ray Fluorescence (EDXRF) and X-ray Fluorescence (XRF) characterized ceramic compositions, while Monte Carlo Particle and Heavy Ion Transport code System (MC PHITS) simulations determined Linear Attenuation Coefficient (LAC), Half-value Layer (HVL), Mass Attenuation Coefficient (MAC), and Mean Free Path (MFP) within the 40-150 kV energy range. Comparative analysis between MC PHITS simulations and real setups was conducted. The C3-S9 ceramic sample, known for homogeneous full-color structure, showcased superior shielding attributes, attributed to its high density and iron content. Notably, energy levels considerably impacted radiation penetration. Overall, C3-S9 demonstrated strong shielding performance, underlining Malaysia's potential ceramic tile resources for X-ray room radiation shielding.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1854-1862
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• 2024

This study aims to investigate the strain-dependent-deformation property of Gyeongju bentonite buffer material. A series of unconfined compressive tests were performed with cylindrical specimens prepared at varying dry densities (𝜌_d = 1.58 g/cm³ to 1.74 g/cm³) using cold isostatic pressing technique. It is found that as 𝜌_d increase, the unconfined compressive strength (q_u), failure strain, and elastic modulus (E) of Gyeongju compacted bentonite (GCB) increases. Normalized elastic modulus (E_sec/E_max) degradation curves of GCB specimens are fitted using Ramberg-Osgood model and the elastic threshold strain (𝜀_e,th) is determined through the fitted curves. The strain-dependency of E and Poisson's ratio (v) of GCB were observed. E and v were measured constant below 𝜀_e,th of 0.14 %. Then, E decreases while v increases after exceeding the strain threshold. The E_sec/E_max degradation curves of GCB in this study suggests wider linear range and higher linearity than those of sedimentary clay in previous study. On top of that, the influence of 𝜌_d is observed on E_sec/E_max degradation curves of GCB, showing a slight increase in 𝜀_e,th with increase in 𝜌_d. Furthermore, an empirical model of q_u with 𝜌_d and a correlation model between q_u and E are proposed for Gyeongju bentonite buffer materials.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2317-2323
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• 2024

Medical linear accelerators with an energy of 8 MV or higher are radiated owing to photonuclear reactions and neutron capture reactions. It is necessary to quantitatively evaluate the concentration of radioactive isotopes when replacing or disposing them. HPGe detectors are commonly used to identify isotopes and measure radioactivity. However, because the detection efficiency is generally calibrated using a standard material with a density of 1.0 g/cm³, a self-absorption effect occurs if the density of the measured material is high. In this study, self-absorption correction factors were calculated for tungsten, lead, copper, and SUS-303, which are the main materials of medical linear accelerator head parts, for each gamma-ray energy using MCNP 6.2 code. The self-absorption effect was more pronounced as the energy of the emitted gamma rays decreased and the density of the measured materials increased. These correction factors were applied to the radioactivity measurements of the in-built and portable HPGe detectors. Furthermore, compared to the surface dose rate measured by the survey meter, the accuracy of the measurements of radioactivity improved by an average of 124.31 and 100.53 % for inbuilt and portable HPGe detectors, respectively. The results showed a good agreement, with an average difference of 3.70 and 5.24 %.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2050-2056
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• 2024

Handling of radioisotopes may cause external and internal contamination to occupational workers while using radiation for medical purposes. This research aims to monitor the internal hazard of occupational workers who handle ¹³¹I. Two methods are used: in vivo or direct method and in vitro or indirect method. The in vivo or direct method was performed by assessing thyroid intake with a thyroid uptake monitoring machine. The in vitro or indirect method was performed by assessing urine samples with the help of a gamma-ray spectroscopy practice using a High-Purity Germanium (HPGe) Detector. In this study, fifty-nine thyroid counts and fifty-nine urine samples were collected from seven occupational workers who were in charge of ¹³¹I at the National Institute of Nuclear Medicine and Allied Sciences (NINMAS), Dhaka. The result showed that the average annual effective dose of seven workforces from thyroid counts were 0.0208 mSv/y, 0.0180 mSv/y, 0.0135 mSv/y, 0.0169 m Sv/y, 0.0072 mSv/y, 0.0181 mSv/y, 0.0164 mSv/y and in urine samples 0.0832 mSv/y, 0.0770 mSv/y, 0.0732 mSv/y, 0.0693 mSv/y, 0.0715 mSv/y, 0.0662 mSv/y, 0.0708 mSv/y.The total annual effective dose (in vivo and in vitro method) was found among seven workers in average 0.1039 mSv/y, 0.0950 mSv/y, 0.0868 mSv/y, 0.0862 mSv/y, 0.0787 mSv/y, 0.0843 mSv/y, 0.0872 mSv/y. Following the rules of the International Commission on Radiological Protection (ICRP), the annual limit of effective dose for occupational exposure is 20 mSv per year and the finding values from this research work are lesser than this safety boundary.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1619-1626
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• 2024

In order to optimize the operational safety and reliability of the upper launcher for the CFETR ECRH system, a design of the launcher for NTM control based on the remote steering concept is currently being carried out for comparison with the front steering equivalent. This paper presents the layout design and analysis of the quasi-optical system in the remote steering launcher. A 3D visual quasi-optical design tool has been developed for the quasi-optical system, which can parameterize modeling, perform general astigmatic beam calculation and show the accurate beam propagation path in the upper port. Three identical sets of quasi-optical modules are arranged in the launcher, and each one consists of two fixed double-curvature focusing mirrors, which focus and reflect the steering beams (- 12°-12°) from two square corrugated waveguides. The beam characteristics at the resonance layer are described, and the average beam radius is < 100 mm. The peak head loads on the surfaces of the two fixed mirrors are 1.63 MW/m² and 1.52 MW/m². The position and size of the beam channel in the blanket are obtained, and the opening apertures on the launcher-facing and plasma-facing sides of the blanket module are 0.54 m² and 0.4 m², respectively.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2718-2731
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• 2024

To ensure the operational safety of nuclear power plants, we present a Quadruped Inspection Robot that can be used for many types of steam generators. Since the Inspection Robot relies on the Holding Modules to grip the tube-sheet, it can be regarded as a hybrid robot with variable configurations, switching between 4-RRR-RR, 3-RRR-RR, and two types of 2-RRR-RR, and the variable configurations bring a great challenge to dimensional synthesis. In this paper, the kinematic model of the Inspection Robot in multiple configurations is established, and the analytical solution is given. The workspace mapping is analyzed by the solution-space, and the workspace of multiple configurations is decomposed into the workspace of 2-RRR to reduce the analysis complexity, and the workspace calculation is simplified by using the envelope rings. The optimization problem of the manipulator is transformed into the calculation of the shortest contraction length of the swing leg. The switching performance of the Inspection Robot is evaluated by stride-length, turning-angle, and workspace overlap-ratio. The performance indexes are classified and transformed based on the proportions and variation trends of dimensional parameters to reduce the number of optimization objective functions, and Pareto optimal solutions are obtained using an intelligent optimization algorithm.