• 기술사
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• 제43권2호
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• pp.45-52
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• 2010

At the end of last year, Korean nuclear power plants, APR-1400 and a research reactor have been contracted to build plants at United Arab Emirates and Jordan. Since 1959, a historical background of nuclear technologies in Korea is summarized. The safety requirements for Instrumentation and Control (I&C) systems in Nuclear Power Plants (NPP) are discussed. Specific descriptions on the typical safety classification of I&C systems, the definitions of the electrical class 1E and the countermeasures against common caused failures are provided. And summaries of typical I&C systems such as the protection systems, the control systems, the instrumentations, the monitoring systems and a control room in NPP are introduced. Strict requirements on the development of the digital computer systems in nuclear applications are described.

• 한국지진공학회논문집
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• 제16권3호
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• pp.35-42
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• 2012

암반지반에 주어진 등재해도 스펙트럼에 상응하는 원전부지 토사지반에서의 등재해도 스펙트럼을 도출하기 위한 확률론적 방법론을 제시하였다. 이를 위해 지진 운동 및 지반의 불확실성을 고려한 지반응답 해석을 통해 토사지반 지표에서의 지진동 증폭계수를 산정하였다. 증폭계수는 가장 상관관계가 높은 지반운동의 스펙트럴 가속도 규모와의 회귀분석을 통해 계산되었다. 이 방법론을 적용하여 국내 KNGR (Korean Next Generation Reactor) 및 APR1400 (Advanced Power Reactor 1400) 원전의 포괄부지 지반 중 B1, B4, C1 및 C3 지반을 대상으로 등재해도 스펙트럼을 도출하였다. 등재해도 스펙트럼을 통해 지진동 발생 빈도 별 위험 주파수 대역을 평가하고 분석하였다. 이 결과는 원전의 종합적 지진리스크 평가 결과를 보다 합리적으로 개선하는 데에 활용될 수 있으며, 향후 다양한 종류의 토사지반에 대한 등재해도 스펙트럼을 평가하는 데에 적용할 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.192-205
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• 2020

Seismic design practices and seismic response analyses of civil structures and nuclear power plants (NPPs) have conventionally used the peak ground acceleration (PGA) or spectral acceleration (S_a) as an intensity measure (IM) of an earthquake. However, there are many other earthquake IMs that were proposed by various researchers. The aim of this study is to investigate the correlation between seismic responses of NPP components and 23 earthquake IMs and identify the best IMs for correlating with damage of NPP structures. Particularly, low- and high-frequency ground motion records are separately accounted in correlation analyses. An advanced power reactor NPP in Korea, APR1400, is selected for numerical analyses where containment and auxiliary buildings are modeled using SAP2000. Floor displacements and accelerations are monitored for the non- and base-isolated NPP structures while shear deformations of the base isolator are additionally monitored for the base-isolated NPP. A series of Pearson's correlation coefficients are calculated to recognize the correlation between each of the 23 earthquake IMs and responses of NPP structures. The numerical results demonstrate that there is a significant difference in the correlation between earthquake IMs and seismic responses of non-isolated NPP structures considering low- and high-frequency ground motion groups. Meanwhile, a trivial discrepancy of the correlation is observed in the case of the base-isolated NPP subjected to the two groups of ground motions. Moreover, a selection of PGA or S_a for seismic response analyses of NPP structures in the high-frequency seismic regions may not be the best option. Additionally, a set of fragility curves are thereafter developed for the base-isolated NPP based on the shear deformation of lead rubber bearing (LRB) with respect to the strongly correlated IMs. The results reveal that the probability of damage to the structure is higher for low-frequency earthquakes compared with that of high-frequency ground motions.

• 대한인간공학회지
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• 제32권1호
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• pp.107-115
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• 2013

Objective: The aim of this study is to identify the training requirements of new nuclear power plant by a comprehensive literature review. Background: The design of instrumental and control systems for New NPP is applied fully digitalized systems. For example, soft-control, large display panels(LDP), and an advanced alarm system were applied to the APR-1400 or SMART. Method: The NUREG-0711 and international guideline of training program was analyzed from the following four phases of SAT(Systemic Approach to Training): Analysis, Design, Development, Implementation and Evaluation. Results: To identify the requirement of training program, 'Feedback' phase was considered and each phase of SAT was classified. Conclusion: A more systematic requirement of training program is needed which considers the computerized system was applied to the new NPP. Application: The results of the publishing can be useful for standardization of the systematic training program for the operators of NPP.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1791-1798
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• 2019

Instrumentation and Control (I&C) systems of nuclear power plants (NPPs) have been continuously digitalized. These systems have a critical role in the operation of nuclear facilities by functioning as the brain of NPPs. In recent years, as cyber security threats to NPP systems have increased, regulatory and policy-related organizations around the world, including the International Atomic Energy Agency (IAEA), Nuclear Regulatory Commission (NRC) and Korea Institute of Nuclear Nonproliferation and Control (KINAC), have emphasized the importance of nuclear cyber security by publishing cyber security guidelines and recommending cyber security requirements for NPP facilities. As described in NRC Regulatory Guide (Reg) 5.71 and KINAC RS015, challenge response authentication should be applied to the critical digital I&C system of NPPs to satisfy the cyber security requirements. There have been no cases in which the most robust response authentication technology like challenge response has been developed and applied to nuclear I&C systems. This paper presents a challenge response authentication mechanism for a Programmable Logic Controller (PLC) system used as a control system in the safety system of the Advanced Power Reactor (APR) 1400 NPP.

• 한국정보통신학회:학술대회논문집
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• 한국정보통신학회 2019년도 춘계학술대회
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• pp.464-466
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• 2019

원자력 발전소에 사용되는 제어기는 높은 신뢰도를 요구한다. 한국형 디지털 원자력 발전소인 APR1400 (Advanced Power Reactor 1400)을 비롯하여, 과거 많은 원자력 발전소에 FPGA (Field Programmable Gate Array)와 CPLD (Complex Programmable Logic Device, 이하 FPGA로 통칭)가 포함된 제어기가 적용되고 있다. 적용 초기에는 FPGA를 일반적인 IC (Integrated Circuit)처럼 기기검증 및 성능시험으로만 검증을 하였다. 이후 90년대에 들어 FPGA검증에 대한 연구가 시작되면서, FPGA가 칩이 되기 전까지를 소프트웨어로 간주하여 IEEE 1012-2004를 적용하여 소프트웨어 확인 및 검증을 하였다. 현재에는 유럽표준인 IEC 62566을 적용하여 많은 검증을 하고 있다. 이 방법은 현재까지 가장 현명한 방법으로 평가 받고 있다. 이유는 기존의 검증 방법에서 문제가 되었던 SoC (System on Chip)의 특징을 검증하는 방법을 충분히 적용하였기 때문이다. 하지만, IEC 62566은 유럽 표준으로 아직 미국에서는 채택을 하지 않고 있으며, FPGA에 대해서는 IEEE 1012를 적용하는 것을 유지하고 있다. IEEE 1012-2004나 IEC 62566은 기술 표준으로 실무에서는 다양한 방법을 적용하여 기술 표준을 충족시켜서 적용하고 있다. 이 논문에서는 SoC의 검증 방법이 적용된 원자력 안전등급 FPGA에 대한 검증 방법의 절차 및 중요사항에 대해 설명하고자 한다.

• 한국건설관리학회논문집
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• 제19권6호
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• pp.34-45
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• 2018

원자력발전소는 총 사업비 약8.6조, 사업기간 10년 이상이 소요되는 대형 국가기반사업이며, 60년 이상 운영되는 국가경제 및 국민의 안전과 밀접한 시설이다. 따라서 설계, 건설, 운영 등 프로젝트 일련의 과정에서 정확하고 경제적 의사결정이 가능하도록 정확한 정보와 일관된 물리적 형상이 유지되어야 한다. 하지만 원전산업은 다른 산업에 비해 복잡하고 규제가 많아 일찌감치 형상관리의 중요성이 널리 인식되어왔지만, 목적과 주체가 명확하지 않아 체계적인 형상관리에 한계가 있었다. 따라서 본 논문은 많은 격실로 이루어진 APR1400을 설계요건과 형상정보를 반영한 실(Room)기반 데이터베이스 구축 방안을 제시하였다. 실 속성을 구분하는 4가지 요소를 선정하였으며, 이를 기반으로 한 데이터베이스를 활용하여 완성도 높고 체계적인 원자력발전소의 형상관리 체계 활용에 일조하고자 한다.

• 시스템엔지니어링학술지
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• 제14권2호
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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.

• 시스템엔지니어링학술지
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• 제16권2호
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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.