• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1616-1622
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• 2022

Selenium has been identified as an element of interest for the safety assessment of a deep geological repository (DGR) for used nuclear fuel. In Canada, groundwaters at DGR depths in sedimentary rocks have been observed to have a high ionic strength. This paper examines the sorption behavior of Se(-II) onto illite, MX-80 bentonite, Queenston shale, and argillaceous limestone in Na-Ca-Cl solutions of varying ionic strength (0.1-6 mol/kgw (m)) and across a pH range of 4-9. Little ionic strength dependence for Se(-II) sorption onto all solids was observed except that sorption at high ionic strength (6 m) was generally slightly lower than sorption at low ionic strength (0.1 m). Illite and MX-80 exhibited the expected results for anion sorption, while shale and limestone exhibited more constant sorption across the pH range tested. A non-electrostatic surface complexation model successfully predicted sorption of Se(-II) onto illite and MX-80 using the formation of an inner-sphere surface complex and an outer-sphere surface complex. Optimized values for the formation reactions of these surface species were proposed.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1003-1015
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• 2022

The new methodologies to derive discharge limits considering operational flexibility according to international safety standards were developed to help reduce the environmental releases of radioactive effluents from nuclear power plants (NPPs). To overcome the limitations of the two existing methods to set up discharge limits assuming a specific statistical distribution of the effluent discharge, two modified equations were newly proposed to directly derive a particular discharge limits corresponding to the target 'compliance probability' based on the actual annual discharge data for a specific NPP and radionuclide groups. By applying these to the actual yearly discharge data of 14 Korean NPPs for 7 radionuclide groups for the past 20 years, the applicability of two new methodologies to actual cases was demonstrated. The 'characteristic value' with approximately a 90% compliance probability for each Korean NPP and radionuclide group was proposed based on the results. The new approaches for setting up the discharge limits and the characteristic values developed in this study are expected to be effectively utilized to foster operator's efforts to progressively reduce the environmental releases of radioactive effluents of NPPs relative to the previous discharge data considering operational flexibilities.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1374-1381
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• 2022

Under the influence of nuclear radiation, the reliability of steam generators (SGs) is an important factor in the efficiency and safety of nuclear power plant (NPP) reactors. Motion planning that remotely manipulates an SG mobile tube-inspection robot to inspect SG heat transfer tubes is the mainstream trend of NPP robot development. To achieve motion planning, conditional traversal is usually used for base position optimization, and then the A^* algorithm is used for path planning. However, the proposed approach requires considerable processing time and has a single expansion during path planning and plan paths with many turns, which decreases the working speed of the robot. Therefore, to reduce the calculation time and improve the efficiency of motion planning, modifications such as the matrix method, improved parent node, turning cost, and improved expanded node were proposed in this study. We also present a comprehensive evaluation index to evaluate the performance of the improved algorithm. We validated the efficiency of the proposed method by planning on a tube sheet with square-type tube arrays and experimenting with Model SG.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3732-3744
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• 2022

To investigate the effects of ocean conditions on the thermal stress and deformation caused by thermal stratification of a pressurizer surge line in a floating nuclear power plant (FNPP), the finite element simulation platform ANSYS Workbench is utilized to conduct the fluid-solid-thermal coupling transient analysis of the surge line under normal "wave-out" condition (no motion) and under ocean conditions (rolling and pitching), generating the transient response characteristics of temperature distribution, thermal stress and thermal deformation inside the surge line. By comparing the calculated results for the three motion conditions, it is found that ocean conditions can significantly improve the thermal stratification phenomenon within the surge line, but may also result in periodic oscillations in the temperature, thermal stress, and thermal deformation of the surge line. Parts of the surge line that are more susceptible to thermal fatigue damage or failure are determined. According to calculation results, the improvements are recommended for pipeline structure to reduce the effects of thermal oscillation caused by ocean conditions. The analysis method used in this study is beneficial for designing and optimizing the pipeline structure of a floating nuclear power plant, as well as for increasing its safety.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2757-2772
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• 2023

The fluid retention effect of the Volume Control Tank (VCT) leads to a long time delay in Reactor Coolant System (RCS) concentration during the boration process. A coupled model combining a lumped-parameter sub-model and a computational fluid dynamics sub-model is currently used to investigate the concentration dynamic characteristic of RCS during the boration process. This model is validated by comparison with experimental data, and the predicted results show excellent agreement with experimental data. We provide detailed fields in VCT and concentration variations of RCS to study the interaction between mixing in VCT and the transient responses of RCS. Moreover, the impacts of the inlet flow rate, inlet nozzle diameter, original concentration, and replenishing temperature of VCT on the RCS concentration characteristic are studied. The inlet flow rate and nozzle diameter of VCT remarkably affect the RCS concentration characteristic. Too-large or too-small inlet flow rates and nozzle diameters will lead to unacceptable long delays. In this work, the optimal inlet flow rate and nozzle diameter of VCT are 5 m³/h and 58.8 mm, respectively. Besides, the impacts of the original concentration and replenishing temperature of VCT are negligible under normal operating conditions.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2897-2904
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• 2023

The Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) method is a widely used method in human reliability analysis (HRA). Performance shaping factors (PSFs) refer to the factors that may influence human performance and are used to adjust nominal human error probabilities (HEPs) in SPAR-H. However, the PSFs are assumed to be independent, which is unrealistic and can lead to unreasonable estimation of HEPs. In this paper, a new method is proposed to handle the dependencies among PSFs in SPAR-H to obtain more reasonable results. Firstly, the dependencies among PSFs are analyzed by using decision-making trial and evaluation laboratory (DEMATEL) method. Then, PSFs are assigned different weights according to their dependent relationships. Finally, multipliers of PSFs are modified based on the relative weights of PSFs. A case study is illustrated that the proposed method is effective in handling the dependent PSFs in SPAR-H, where the duplicate calculations of the dependent part can be reduced. The proposed method can deal with a more general situation that PSFs are dependent, and can provide more reasonable results.

• Journal of Korean Association for Spatial Structures
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• v.23 no.3
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• pp.35-43
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• 2023

This study aims to evaluate the structural safety of a structural thermal barrier, installed inside the structure of a building and performed the role of a load-bearing element and an insulation simultaneously, contributing to the realization of net-zero buildings. To ensure the reliability of the analysis model, the analysis results derived from LS-DYNA were compared with the experimental results. Based on the results shown through the flexural experiment, the reliability of the thermal cross-section insulation structure model for slabs was validated. In addition, the effect of the UHPC block on the load support performance and its contribution to vertical deflection was verified.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3506-3513
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• 2023

In this paper, we demonstrate the neutron attenuation of dry cask shielding materials using the S670e helium-4 detector manufactured by Arktis Radiation Ltd. In particular, two materials expected to be applied to the TN-32 dry cask manufactured by ORANO Korea and KORAD-21 by the Korea Radioactive Waste Agency (KORAD) were utilized. The measured neutron attenuation was compared with our Monte Carlo N-Particle Transport simulation results, and the difference is given as the root mean square (RMS). For the fast neutron case, a rapid decline in neutron counts was observed as a function of increasing material thickness, exhibiting an exponential relationship. The discrepancy between the experimentally acquired data and simulation results for the fast neutron was maintained within a 2.3% RMS. In contrast, the observed thermal neutron count demonstrated an initial rise, attained a maximum value, and exhibited an exponential decline as a function of increasing thickness. In particular, the discrepancy between the measured and simulated peak locations for thermal neutrons displayed an RMS deviation of approximately 17.3-22.4%. Finally, the results suggest that a minimum thickness of 5 cm for Li-6 is necessary to achieve a sufficiently significant cross-section, effectively capturing incoming thermal neutrons within the dry cask.