• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.649-655
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• 2013

During welding, the dissimilar metal butt welds of nuclear piping are typically subjected to repair welding in order to eliminate defects that are found during post-weld inspection. It has been found that the repair weld can significantly increase the tensile residual stress in the weldment, and therefore, accurate estimation of the weld residual stress due to repair weld, especially for dissimilar metal welds using Ni-based alloy 82/182 in nuclear components, is of great importance in order to assess susceptibility to primary water stress corrosion cracking. In the present study, the stress distributions of dissimilar metal butt welds in nuclear reactor piping subjected to repair weld were investigated based on detailed nonlinear finite element analyses. Particular emphasis was placed on the variation of the stress distribution in the dissimilar metal butt weld according to the finite element welding analysis sequence for the repair welding process.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.9
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• pp.2894-2900
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• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.

• Journal of KIISE:Software and Applications
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• v.32 no.12
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• pp.1178-1191
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• 2005

In this paper, we propose a synthesis method of software fault tree from software requirements specification written in NuSCR formal specification language. The software fault tree, proposed in this paper, reflects requirements on both structure and behavior and it is an integrated form. The software fault tree can be used for analyzing safety in the view of structure and behavior. We propose templates for each components in NuSCR specification language and a synthesis method of software fault tree using the templates. The research was applied into the main trip logic of the reactor protection system of ARP1400, the Korean next generation nuclear reactor system, developed by KNICS. And we evaluate feasibility of our approach through this case study.

• Journal of Energy Engineering
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• v.26 no.3
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• pp.71-77
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• 2017

In the event of loss of coolant accident (LOCA) and station black out (SBO) in the primary system of a nuclear reactor, the coolant water should be injected to reactor coolant system (RCS) without any intervention of operators or active components. To satisfy the requirements, hybrid safety injection tank (Hybrid SIT) was suggested by Korea Atomic Energy Research Institute (KAERI). The pressure equalizing time of Hybrid SIT is an important parameter to determine the timing of coolant injection. To predict the pressure equalizing time of the Hybrid SIT, a separate effect test facility was constructed and sensitivity tests were conducted in various conditions. The main parameter determining the pressure equalizing time was obtained from separate effect test (SET) results. The wall of condensation on the inner wall of SIT and direct contact condensation on the water surface affected to the pressure equalizing time very much. In this study, the effect of each condensation phenomena on pressure equalizing time was quantitatively analyzed from results of SET and a prediction method of pressure equalizing time was proposed.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.613-624
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• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• The KSFM Journal of Fluid Machinery
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• v.16 no.4
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• pp.35-43
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• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.37-50
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• 2020

A passive containment cooling system (PCCS) has been developed as advanced safety feature for innovative power reactor (iPOWER). Passive systems are inherently less stable than active systems and the PCCS encountered the flashing-induced instability previously identified. The objective of this study is to develop stability maps for flashing-induced instability using MARS (Multi-dimensional Analysis of Reactor Safety) code. Firstly, we conducted a series of sensitivity analysis to see the effects of time step size, nodalization, and alternative MARS user options on the onset of flashing-induced instability. The riser nodalization strongly affects the prediction of flashing in a long riser of the PCCS, while time step size and alternative user options do not. Based on the sensitivity analysis, a standard input and an analysis methodology were set up to develop the stability maps of PCCS. We found out that the calculated equilibrium quality at the exit of the riser as a stability boundary above 5 kW/㎡ was approximately 1.2%, which was in good agreement with Furuya's results. However, in case of a very low heat flux condition, the onset of instability occurred at the lower equilibrium quality. In addition, it was confirmed that inlet throttling reduces the unstable region.

• Fire Science and Engineering
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• v.32 no.6
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• pp.141-149
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• 2018

The perception of risk that firefighters have is closely related to their performance and emergency preparedness in nuclear power plant accidents. This study investigated the unique risk perception among firefighters working in nuclear power plants (NPPs) using a concept mapping method. Thirty three firefighters in NPPs participated in this study. Two core axes, "fear and control" and "coping resource", emerged in the firefighters' risk perception. In particular, the risk perception consisted of six clusters: fear of radiation exposure and low controllability; anxiety caused by the lack of control and authority; lack of trust and cooperation; lack of authority and professionals; lack of equipment, manual, and information; and lack of knowledge and training. Catastrophic expectation and a low sense of control caused by the lack of responsive resources were the main factors that increase the risk perception. The theoretical and practical contributions of this study were discussed.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.589-600
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• 2022

Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particles conducted at the Sun Yat-sen University. To better visualize and clarify the mechanism and characteristics of sloshing induced by local Fuel-Coolant Interaction (FCI), experiments were performed with various parameters by injecting nitrogen gas into a 2-dimensional liquid pool with accumulated solid particles. It was confirmed that under different particle-bed conditions, three representative flow regimes (i.e. the bubble-impulsion dominant, transitional and bed-inertia dominant regimes) are identifiable. Aimed at predicting the regime transitions during sloshing process, a predictive empirical model along with a regime map was proposed on the basis of experiments using single-sized spherical solid particles, and then was extended for covering more complex particle conditions (e.g. non-spherical, mixed-sized and mixed-density spherical particle conditions). To obtain more comprehensive understandings and verify the applicability and reliability of the predictive model under more realistic conditions (e.g. large-scale 3-dimensional condition), further experimental and modeling studies are also being prepared under other more complicated actual conditions.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3409-3416
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• 2023

Licensing the next-generation of nuclear reactor designs requires extensive use of Modeling and Simulation (M&S) to investigate system response to many operational conditions, identify possible accidental scenarios and predict their evolution to undesirable consequences that are to be prevented or mitigated via the deployment of adequate safety barriers. Deep Learning (DL) and Artificial Intelligence (AI) can support M&S computationally by providing surrogates of the complex multi-physics high-fidelity models used for design. However, DL and AI are, generally, low-fidelity 'black-box' models that do not assure any structure based on physical laws and constraints, and may, thus, lack interpretability and accuracy of the results. This poses limitations on their credibility and doubts about their adoption for the safety assessment and licensing of novel reactor designs. In this regard, Physics Informed Neural Networks (PINNs) are receiving growing attention for their ability to integrate fundamental physics laws and domain knowledge in the neural networks, thus assuring credible generalization capabilities and credible predictions. This paper presents the use of PINNs as surrogate models for accidental scenarios simulation in Nuclear Power Plants (NPPs). A case study of a Loss of Heat Sink (LOHS) accidental scenario in a Nuclear Battery (NB), a unique class of transportable, plug-and-play microreactors, is considered. A PINN is developed and compared with a Deep Neural Network (DNN). The results show the advantages of PINNs in providing accurate solutions, avoiding overfitting, underfitting and intrinsically ensuring physics-consistent results.