• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.47-58
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• 2020

Most of the reactor trips in Korean NPPs related to core protection systems were caused not because of proximity of boiling crisis and, consequently, a damage in the core, but due to particular miscalculations or component failures related to the core protection system. The most common core protection system applied in Korean NPPs is the Core Protection Calculator System (CPCS), which is installed in OPR1000 and APR1400 plants. It generates a trip signal to scram the reactor in case of low Departure from Nucleate Boiling Ratio (DNBR) or high Local Power Density (LPD). However, is a reactor trip necessary to protect the core? Or could a fast power reduction be enough to recover the DNBR/LPD without a scram? In order to analyze the online calculation of DNBR/LPD, and the use of fast power reduction as trip avoidance methodology, a concept of CPCS with fast power reduction function was developed in Matlab® Simulink using systems engineering approach. The system was validated with maximum of 0.2% deviation from the reference and the dynamic deviation was maximum of 12.65% for DNBR and 6.72% for LPD during a transient of 16,000 seconds.

• Proceedings of the KIEE Conference
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• 2000.07d
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• pp.2498-2500
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• 2000

차세대 신형원전에서는 디지털 기술의 적용을 기본 설계 요건으로 제시하고 있다. 차세대 원전의 노심보호계통 (Core Protection Calculator Systems; CPCS)은 원전의 안전성을 보장하기 위한 부분으로 이 부분이 올바르게 작성되고, 검증되어야 함은 분명하다. 현재 이부분은 소프트웨어로 개발 중에 있으며 개발 단계에 있어서 시작단계인 요구명세 단계에 있다. 요구 명세 단계의 오류는 소프트웨어 개발 단계 중 소프트웨어의 품질에 가장 영향을 많이 미치는 단계로 알려져 있으므로 이 단계를 정확하게 수행하여야 한다. 안전성이 중요한 소프트웨어를 명세하는 데 있어서 우선 정의되어야 하는 것은 어떤 절차를 통해서 어떤 방법으로 할지를 결정하여 그 절차를 정하여야 한다. 기존에 소프트웨어 요구 명세에 대한 표준안이 존재하기는 하지만, 이러한 표준안들은 개념적인 언어들로 쓰여져 있기 때문에 실제 소프트웨어의 개발 과정에 사용하기 위해서는 구체적인 언어들로 다시 작성하여야 한다. 따라서, 소프트웨어 명세를 작성하기 위해서 절차와 방법에 대해서 정의하여야 한다. 본 논문에서는 개략적인 명세 절차와 명세 방법등을 기술하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.99.1-99
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• 2001

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.41-48
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• 2018

This paper describes the systems engineering development process for the Departure from Nucleate Boiling Ratio (DNBR) algorithm using FPGA. Current Core Protection Calculator System (CPCS) requirement and DNBR logic are analyzed in the reverse engineering phase and the new FPGA based DNBR algorithm is designed in the re-engineering phase. FPGA based DNBR algorithm is developed by VHSIC Hardware Description Language (VHDL) in the implementation phase and VHDL DNBR software is verified in the software Verification & Validation phase. Test cases are developed to perform the software module test for VHDL software modules. The APR 1400 simulator is used to collect the inputs data in 100%, 75%, and 50% reactor power condition. Test input signals are injected to the software modules following test case tables and output signals are compared with the expected test value. Minimum DNBR value from developed DNBR algorithm is validated by KEPCO E&C CPCS development facility. This paper summarizes the process to develop the FPGA-based DNBR calculation algorithm using systems engineering approach.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.24-32
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• 2018

System Complexity one of the most common cause failure of the projects, it leads to a lack of understanding about the functions of the system. Hence, the model is developed for communication and furthermore modeling help analysis, design, and understanding of the system. On the other hand, the text-based specification is useful and easy to develop but is difficult to visualize the physical composition, structure, and behaviour or data exchange of the system. Therefore, it is necessary to transform system description into a diagram which clearly depicts the behaviour of the system as well as the interaction between components. According to the International Atomic Energy Agency (IAEA) Safety Glossary, The safety system is a system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents. Core Protection Calculator System (CPCS) in Advanced Power Reactor 1400 (APR 1400) Nuclear Power Plant is a safety critical system. CPCS was developed using systems engineering method focusing on Departure from Nuclear Boiling Ratio (DNBR) calculation. Due to the complexity of the system, many diagrams are needed to minimize the risk of ambiguities and lack of understanding. Using Model-Based Systems Engineering (MBSE) software for modeling the DNBR algorithm were used. These diagrams then serve as the baseline of the reverse engineering process and speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.45-57
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• 1990

A steady-state margin comparison study was performed between analog and digital protection systems. The systems compared are the thermal overpower and overtemperature delta T system of Westinghouse, and Core Protection Calculator System of Combustion Engineering, Inc. No dynamic offset was considered to eliminate the margin differences by different safety analysis methodologies. The result shows that the digital protection system has about 30% more rated power margin than the analog system in protecting against the fuel rod centerline melting. The digital protection system is shown to have almost same margin with the analog protection system in preventing the DNB at EOC (End of Cycle) even if the digital protection system has about 10% more margin at BOC(Beginning of Cycle).

• Journal of Institute of Control, Robotics and Systems
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• v.4 no.3
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• pp.400-405
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• 1998

본 논문에서는 한국형 원자력 발전소의 기준모델인 영광 3,4호기 운전원 훈련용 시뮬레이터의 모델링 절차와 ABB-CE 원전의 독특한 계통인 CPC/COLSS (Core protection Calculator/Core Operating Limit Supervisory System) 계통에 대한 모델링을 전개허고 있다. CPC/COLSS는 원자로를 포함하는 냉각재계통(NSSS)과 핵연료의 건전성을 보장하기위한 계통으로서 감시및 보호 과정에서의 계산을 디지털화시킴으로서 정확성과 함께 원자로의 안정성을 향상시킨 특색있는 계통이다. 따라서 영광 3,4호기 시뮬레이터에서는 CPC/COLSS 계통에 대한 정확한 모델링을 하여 시험을 통해 성능및 기능에 대한 검증을 마침으로서 CPC/COLSS 시뮬레이션 모델 개발이 성공적으로 되었고 영광 3,4호기 운전 특성에 맞는 시뮬레이터를 개발하였다.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.2
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• pp.49-60
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• 2021

The Core Protection Calculator System (CPCS) is vital for plant safety as it ensures the required Specified Acceptance Fuel Design Limit (SAFDL) are not exceeded. The CPCS generates trip signals when Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) exceeds their predetermined setpoints. These setpoints are established based on the operating margin from the analysis that produces the SAFDL values. The goal of this research is to create a simplified CPCS that optimizes operating margin for I-SMRs. Because the I-SMR is compact in design, instrumentation placement is a challenge, as it is with Ex-core detectors and RCP instrumentation. The proposed CPCS addresses the issue of power flux measurement with In-Core Instrumentation (ICI), while flow measurement is handled with differential pressure transmitters between Steam Generators (SG). Simplification of CPCS is based on a Look-Up-Table (LUT) for determining the CEA groups' position. However, simplification brings approximations that result in a loss of operational margin, which necessitates compensation. Appropriate compensation is performed based on the result of analysis. FPGAs (Field Programmable Gate Arrays) are presented as a way to compensate for the inadequacies of current systems by providing faster execution speeds and a lower Common Cause Failure rate (CCF).

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.877-887
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• 1995

YGN 3 is the first nuclear power plant to use the Core Protection Calculator (CPC) as the core protection system and the Core Operating Limit Supervisory System (COLSS) as the core monitor-ing system in Korea. The CPC is designed to provide on-line calculations of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) and to initiate reactor trip if the core conditions exceed the DNBR or LPD design limit. The COLSS is designed to assist the operator in implementing the Limiting Conditions for Operation (LCOs) in Technical Specifications for DNBR/Linear Heat Rate (LHR) margin, azimuthal tilt, and axial shape index and to provide alarm when the LCOs are reached. During YGN 3 initial startup testing, extensive CPC/COLSS related tests ore peformed to ver-ify the CPC/COLSS performance and to obtain optimum CPC/COLSS calibration constants at var, -ious core conditions. Most of test results met their specific acceptance criteria. In the case of missing the acceptance criteria, the test results ore analyzed, evaluated, and justified. Through the analysis and evaluation of each of the CPC/COLSS related test results, it can be concluded that the CPC/COLSS are successfully Implemented as designed at YGN 3.

• The Transactions of the Korean Institute of Electrical Engineers D
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• v.55 no.7
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• pp.303-312
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• 2006

This paper proposes the architecture and protocol of a data communication network for the safety system in nuclear power plants. First, we establish four design criteria with respect to determinability, reliability, separation and isolation, and verification/validation. Next we construct the architecture of the safety network for the following systems: PPS (Plant Protection System), ESF-CCS (Engineered Safety Features-Component Control System) and CPCS (Core Protection Calculator System). The safety network consists of 12 sub-networks and takes the form of a hierarchical star. Among 163 communication nodes are about 1600 origin-destination (OD) pairs created on their traffic demands. The OD pairs are allowed to exchange data only during the pre-assigned time slots. Finally, the communication protocol is designed in consideration of design factors for the safety network. The design factors include a network topology of star, fiber-optic transmission media, synchronous data transfer mode, point-to-point link configuration, and a periodic transmission schedule etc. The resulting protocol is the modification of IEEE 802.15.4 (LR-WPAN) MAC combined with IEEE 802.3 (Fast Ethernet) PHY. The MAC layer of IEEE 802.15.4 is simplified by eliminating some unnecessary (unctions. Most importantly, the optional TDMA-like scheme called the guaranteed time slot (GTS) is changed to be mandatory to guarantee the periodic data transfer. The proposed protocol is formally specified using the SDL. By performing simulations and validations using Telelogic Tau SDL Suite, we find that the proposed safety protocol fits well with the characteristics and the requirements of the safety system in nuclear power plants.