• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1638-1653
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• 2024

Given the unpredictability of the occurrence of the earthquake and other potential disasters into consideration, the nuclear power plant may be confronted with beyond design-basis earthquake load in the future. The containment structure may be severely damaged under such severe earthquake loading, increasing the risk of containment concrete cracking and potential radioactive materials leaking. Moreover, initial damage caused by the earthquake may significantly alter the pressure performance of the containment under follow-up internal pressure. To compromise the dangers of beyond design-basis earthquake to the containment, an alternative of replacing the conventional concrete with fiber-reinforced concrete (FRC) to upgrade the seismic resistance capacity of the containment is attempted and thoroughly researched. In this study, the influence of various fiber types such as rigid fiber and mixed fiber is regarded to constitute fiber-reinforced PCCVs. The physical properties of traditional and fiber-reinforced PCCVs under earthquake ground motions are scientifically compared and identified by using traditional and proposed evaluation indices. The results indicate that both the traditional evaluation index (i.e. top displacement, stress, strain) and the proposed damage index are greatly reduced by the practice of fiber strengthening under earthquake ground motions.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3826-3834
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• 2024

The main purposes of start-up tests in nuclear power plants (NPPs) are to ensure safe and reliable operation, verify system functionality, comply with regulatory requirements, optimize performance, and establish a foundation for ongoing plant operation and maintenance. However, the start-up tests of NPPs also could be used as a main part of the nuclear code validation procedure for several reasons including: realistic simulation, comprehensive evaluation, detection of code limitations, validation of safety margins and confidence in code predictions. The main purpose of the current study is to define and assess the validation procedure based on actual start-up test data. In this regard, the developed RELAP5 model has been validated against the actual data of VVER-1000 plant during a dynamic start-up test. The results of this full-scale validation show a good agreement between the developed RELAP5 model results and actual plant data. Finally, by defining a step by step validation procedure, it has been recommended to use the start-up phase test data as a more robust validation process which allow for full-scale validation of the nuclear code by comparing its predictions with actual plant measurements and also other advantages which have been demonstrated in the current study.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3851-3863
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• 2024

The speed of offsite consequence analysis is highly important due to the extensive calculations required to handle all the scenarios for a single-unit or multi-unit Level 3 PSA (probabilistic safety assessment). To perform an offsite consequence analysis as part of Level 3 PSA, various input parameters are considered, amongst which certain parameters, such as plume segments, spatial grids, and particle size distributions, have flexible input formats. This study describes the development of an effective optimization method to reduce the analysis time as much as possible while maintaining the accuracy of the offsite consequence analysis results. The effect of plume segmentation on offsite consequence analysis was investigated by observing deviations in analysis results and changes in the required analysis times following changes in plume release. Then a plume segmentation optimization method based on the cumulative release fraction slope was developed to intensively analyze the sections with rapid release and to simplify the analysis for the sections with nonsignificant release. As a result of applying this method, the analysis time was reduced by about 54.5 % compared to the base case, while the resulting health effects showed very small deviations of 0.03 % and 1.77 % for early fatality risk and cancer fatality risk, respectively.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3450-3462
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• 2024

This paper presents a computational framework for drop time assessment of a control element assembly (CEA) under fuel assembly (FA) deformations. The proposed framework consists of three key components: 1) finite element modeling of CEA, 2) fluid-structure interaction to compute drag force, and 3) modeling of frictional contact between CEA and FA. Specially, to accommodate the large motion of CEA, beam elements based on absolute nodal coordinate formulation (ANCF) are adopted. The continuity equation is utilized to calculate the drag force, considering flow changes in the cross-sectional area during the CEA drop. Lastly, beam-inside-beam frictional contact model is employed to capture practical contact conditions between CEA and FA. The proposed framework is validated through experiments under two scenarios: free falls of CEA within FA, encompassing undeformed and deformed scenarios. The experimental validation of the framework demonstrated that the drop time of CEA can be accurately predicted under the complex coupling effects of fluid and frictional contact. The drop times of the S-shaped deformation case is longer than those of the C-shaped deformation case, affirming the time delay due to frictional force. The validation confirms the potential applicability to access the safety and reliability of nuclear power plants under extreme conditions.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3330-3334
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• 2024

Void drift between subchannels in a rod bundle is a crucial phenomenon affecting the calculation accuracy of thermal-hydraulic parameters in SMRs. It holds significant importance in enhancing the precision of safety analysis for SMRs. Existing research on experiment and model of void drift between rod bundle subchannels is relatively rare, and the accuracy of model calculations requires improvement. In this study, experiments on gas-liquid two-phase non-equilibrium flow were conducted to measure the redistribution of two-phase flow induced by void drift in a 1 × 2 rod bundle. The experiment results indicated that in bubby flow regime with void fraction less than 0.3, the void diffusion coefficient showed little variation with changes in void fraction. However, in slug flow and annular flow regimes with void fraction exceeding 0.3, the void diffusion coefficient significantly increased with an increase in void fraction. Furthermore, a new void drift model was developed and validated based on a subchannel code. The overall predicted uncertainty for the outlet void fraction in the rod bundle benchmark was less than 13%.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3167-3179
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• 2024

In abnormal states of nuclear power plants (NPPs), operators undertake mitigation actions to restore a normal state and prevent reactor trips. However, in abnormal states, the NPP condition fluctuates rapidly, which can lead to human error. If human error occurs, the condition of an NPP can deteriorate, leading to reactor trips. Sudden shutdowns, such as reactor trips, can result in the failure of numerous NPP facilities and economic losses. This study develops a remaining trip time (RTT) prediction system as part of an operator support system to reduce possible human errors and improve the safety of NPPs. The RTT prediction system consists of an algorithm that utilizes artificial intelligence (AI) and explainable AI (XAI) methods, such as autoencoders, light gradient-boosting machines, and Shapley additive explanations. AI methods provide diagnostic information about the abnormal states that occur and predict the remaining time until a reactor trip occurs. The XAI method improves the reliability of AI by providing a rationale for RTT prediction results and information on the main variables of the status of NPPs. The RTT prediction system includes an interface that can effectively provide the results of the system.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3512-3527
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• 2024

Seismic quantification of nonstructural components like electrical cabinets is essential to ensure the uninterrupted operation of nuclear facilities during earthquake events. This process requires experimental tests, which can be expensive, time-consuming, and limited by safety concerns and precision. As an alternative to that, numerical simulations should be done in such a way that they are capable of capturing the global dynamic behavior with minimum computational efforts. However, in the case of complex interconnected cabinets, the simplification of numerical models often poses difficulties in illustrating the real-time behavior of combined cabinet systems. On the other hand, detailed three-dimensional (3D) numerical models require lengthy time and sophisticated computational setup, indicating their expensive computational efforts. To resolve this issue, a simplified and efficient 3D modeling approach has been proposed in this study. The accuracy of the results from the new model showed an excellent match with that obtained from the responses of the experimental test. After the validation and observation of the numerical efficiency, a practical application is implemented by considering the impacts of earthquake frequency contents on the behavior of cabinet systems. From the outcomes, it is evident that this proposed modeling methodology has the potential to replace the complex combined nuclear cabinet models for earthquake evaluation.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3626-3643
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• 2024

One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for validation of the reactor core design code and its uncertainty quantification (UQ). This study aims to quantify the criticality uncertainty of a typical TRU burner with metallic fuel by using the standard upper safety limit (USL) estimation framework based on the similarity analysis of existing benchmark experiments but elaborated in two aspects:1) application of two-sided rather than one-sided tolerance interval and 2) inclusion of additional uncertainty to account for fission products and minor actinides not included in the benchmark experiments. To conduct the similarity analysis and evaluate the nuclear-data induced uncertainty, existing, well-verified computing codes were integrated, including the nuclear data sampling code SANDY, the nuclear data processing code NJOY, and the continuous-energy Monte Carlo code McCARD. Finally, using the SFR benchmark database comprising both publicly available and proprietary benchmark experiments, the criticality uncertainty of the TRU core model with metallic fuel was evaluated.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3812-3825
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• 2024

The axial-flow-induced vibration of fuel rods in the nuclear power plant is closely related to nuclear safety. In this article, a numerical study is performed on vibration of two elastic cylinders arranged side-by-side in axial flow. Large eddy simulation is employed to predict the turbulent flow. The numerical method has been verified using the experimental root-mean-square vibration amplitude of a single cylinder. A wide range of inflow velocities u^*, incident turbulence intensity T_u and space ratio P/D have been examined, where D and P are the diameter and centre-to-centre distance of the cylinders, respectively. The results show that the vibration amplitudes increase with an increasing u^*, comparable to the case of a single cylinder in axial flow. However, the two cylinders could bend outwards during a relatively high u^* and low T_u. Although T_u significantly affects the amplitudes of the cylinders, it does not change the vibration frequency and the critical velocity at which buckling instability occurs. As the gap between the two cylinders is sufficiently small, the vibration amplitude enhances significantly due to the pronounced hydrodynamic interaction between the two elastic cylinders and surrounding fluid. The direction of buckling is no longer random but fixed.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3144-3154
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• 2024

The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive and Discrete Gadolinia/Alumina Burnable Absorber (HIGA) rods for controlling excess reactivity sustainably. Through comprehensive analysis and simulations, the reactivity behavior with varying quantities of HIGA rods is examined, leading to the development of optimized fuel assembly designs. Furthermore, the integration of HIGA rods with integral gadolinia BA rods is discussed to enhance reactivity control and operational flexibility further. This approach utilizes the spatial self-shielding effect of gadolinia for extended reactivity management, crucial for stable and efficient reactor performance. The paper thoroughly addresses core design considerations, including fuel assembly configurations and control rod patterns, to ensure safety and performance in initial and reload cycles. This research advances the development of SBF operation in i-SMR by offering practical reactivity management solutions.