• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.1-15
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• 2020

Nuclear power plants in South Korea are operated to cover the baseload demand. Hence they are operated at 100% rated power and do not deploy power tracking control except for startup, shutdown, or during transients. However, as the contribution of renewable energy in the energy mix increases, load follow operation may be needed to cover the imbalance between consumption and production due to the intermittent nature of electricity produced from the conversion of wind or solar energy. Load follow operation may be quite challenging since the operators need to control the axial power distribution and core reactivity while simultaneously conducting the power maneuvering. In this paper, a systems engineering approach for multi-physics load follow simulation of APR1400 is performed. RELAP5/SCDAPSIM/MOD3.4/3DKIN multi-physics package is selected to simulate the Korean Advanced Power Reactor, APR1400, under load follow operation to reflect the impact of feedback signals on the system safety parameters. Furthermore, the systems engineering approach is adopted to identify the requirements, functions, and physical architecture to provide a set of verification and validation activities that guide this project development by linking each requirement to a validation or verification test with predefined success criteria.

• The Transactions of the Korean Institute of Power Electronics
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• v.27 no.5
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• pp.425-430
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• 2022

A pumped storage power station is an important means to solve the problem of peak load regulation and ensures the safety of power grid operation. The doubly fed variable-speed pumped storage (DFVSPS) system adopts a doubly fed induction machine (DFIM) to replace the synchronous machine used in traditional pumped storage. The stator of DFIM is connected to the power grid, and the three-phase excitation windings are symmetrically distributed on the rotor. Excitation current is supplied by the converter. The active and reactive power of the unit can be quickly adjusted by adjusting the amplitude, frequency, and phase of the rotor-side voltage or current through the converter. Compared with a conventional pumped storage hydropower station (C-PSH), DFVSPS power stations have various operating modes and frequent start-up and shutdown. This study introduces the structure and principle of the DFVSPS unit. Mathematical models of the unit, including a model of DFIM, a model of the pump-turbine, and a model of the converter and its control, are established. Fast power control strategies are proposed for the unit model. A 300 MW model of the DFVSPS unit is established in MATLAB/Simulink, and the response characteristics in generating mode are examined.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.907-915
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• 2024

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

• Korean Chemical Engineering Research
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• v.58 no.4
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• pp.604-609
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• 2020

The purpose of this study is to check the priority control status of the current operation process alarm by comparing the priority of the alarm set up in PV high trip point, which is being installed and operated in the domestic petrochemical industry, with the criteria presented in ISA 18.2 International Standard or EEMUA 191 Guidelines. In the event of a process problem, excessively set high alarm is provided to the driver in a short period of time, making it difficult to identify the alarm that needs to be handled first. As a result, it is likely that the operator will not be able to carry out appropriate actions within the specified time frame, and many cases have been reported leading to unexpected process shutdowns or process accidents. Therefore, this study aims to introduce international standards related to alarm management and identify the level of alarm control used at the domestic petrochemical industry site to inform potential risks that may occur in the petrochemical process of the national industrial complex in the future and suggest ways to reduce risk factors. This paper was submitted to Professor Lee Inbeom's retirement anniversary issue.

• Fire Science and Engineering
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• v.26 no.6
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• pp.38-44
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• 2012

The purpose of fire protection for the nuclear power plants (NPPs) is to ensure safe shutdown state of the reactor, to minimize the release of radioactive materials to the environment, to provide physical safety of the on-site personnel, and to limit the property damage. Fire protection and extinguishing equipments are one of the important protection measures based on the defense-in-depth concept, which can promptly detect and control and extinguish those fires that do occur, thereby limiting fire damage. However, a separate evaluation process might be additionally necessary for the construction permit and operating license because the fire protection laws of the NEMA for installation standards of the fire protection systems is not fully characterized for the NPPs. It is also not easy to implement the regulations such as the performance based design concept for fire protection system of the NPPs which are characterized for a relatively low density of employee. This study suggests a guideline for the improvement of the technical standards for fire protection systems of the NPPs by evaluating the fundamental problems drawn by reviewing laws and regulatory guides relevant to fire protection and by evaluating the applicability of the KEPIC FPN in domestic nuclear power plants.

• Journal of KIISE:Software and Applications
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• v.29 no.12
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• pp.860-872
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• 2002

Fault tree analysis is the most widely used saftly analysis technique in industry. However, the analysis is often applied manually, and there is no systematic and automated approach available to validate the analysis result. In this paper, we demonstrate that a real-time model checker UPPAAL is useful in formally specifying the required behavior of safety-critical software and to validate the accuracy of manually constructed fault trees. Functional requirements for emergency shutdown software for a nuclear power plant, named Wolsung SDS2, are used as an example. Fault trees were initially developed by a group of graduate students who possess detailed knowledge of Wolsung SDS2 and are familiar with safety analysis techniques including fault tree analysis. Functional requirements were manually translated in timed automata format accepted by UPPAAL, and the model checking was applied using property specifications to evaluate the correctness of the fault trees. Our application demonstrated that UPPAAL was able to detect subtle flaws or ambiguities present in fault trees. Therefore, we conclude that the proposed approach is useful in augmenting fault tree analysis.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1127-1132
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• 2013

The failure of reactor internals may have a significant effect on the safe operation and shutdown of a reactor. Various agings related to neutron irradiation occur or can potentially occur in the reactor internals owing to high neutron irradiation levels. Austenitic stainless steel, one of the principal materials constituting the reactor internals, shows different mechanical material behaviors such as tensile/creep properties and fracture toughness with neutron irradiation levels. This variation should be considered when the structural integrity of the reactor internals against agings during the design lifetime or continued operation period is evaluated. In this study, user subroutine programs considering the variation of mechanical material behaviors with neutron irradiation levels were developed. The programs were validated by testing them for various conditions.

• Journal of Auto-vehicle Safety Association
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• v.11 no.2
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• pp.35-43
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• 2019

This paper describes the improved environment recognition algorithms using some type of sensors like LiDAR and cameras. Additionally, integrated control algorithm for an autonomous vehicle is included. The integrated algorithm was based on C++ environment and supported the stability of the whole driving control algorithms. As to the improved vision algorithms, lane tracing and traffic sign recognition were mainly operated with three cameras. There are two algorithms developed for lane tracing, Improved Lane Tracing (ILT) and Histogram Extension (HIX). Two independent algorithms were combined into one algorithm - Enhanced Lane Tracing with Histogram Extension (ELIX). As for the enhanced traffic sign recognition algorithm, integrated Mutual Validation Procedure (MVP) by using three algorithms - Cascade, Reinforced DSIFT SVM and YOLO was developed. Comparing to the results for those, it is convincing that the precision of traffic sign recognition is substantially increased. With the LiDAR sensor, static and dynamic obstacle detection and obstacle avoidance algorithms were focused. Therefore, improved environment recognition algorithms, which are higher accuracy and faster processing speed than ones of the previous algorithms, were proposed. Moreover, by optimizing with integrated control algorithm, the memory issue of irregular system shutdown was prevented. Therefore, the maneuvering stability of the autonomous vehicle in severe environment were enhanced.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3879-3891
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• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.