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

• Journal of the Korean Society of Systems Engineering
• /
• v.14 no.2
• /
• pp.73-82
• /
• 2018

In this work, a hardware based cryptographic module for the cyber security of nuclear power plant is developed using a system engineering approach. Nuclear power plants are isolated from the Internet, but as shown in the case of Iran, Man-in-the-middle attacks (MITM) could be a threat to the safety of the nuclear facilities. This FPGA-based module does not have an operating system and it provides protection as a firewall and mitigates the cyber threats. The encryption equipment consists of an encryption module, a decryption module, and interfaces for communication between modules and systems. The Advanced Encryption Standard (AES)-128, which is formally approved as top level by U.S. National Security Agency for cryptographic algorithms, is adopted. The development of the cyber security module is implemented in two main phases: reverse engineering and re-engineering. In the reverse engineering phase, the cyber security plan and system requirements are analyzed, and the AES algorithm is decomposed into functional units. In the re-engineering phase, we model the logical architecture using Vitech CORE9 software and simulate it with the Enhanced Functional Flow Block Diagram (EFFBD), which confirms the performance improvements of the hardware-based cryptographic module as compared to software based cryptography. Following this, the Hardware description language (HDL) code is developed and tested to verify the integrity of the code. Then, the developed code is implemented on the FPGA and connected to the personal computer through Recommended Standard (RS)-232 communication to perform validation of the developed component. For the future work, the developed FPGA based encryption equipment will be verified and validated in its expected operating environment by connecting it to the Advanced power reactor (APR)-1400 simulator.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.20 no.1
• /
• pp.25-31
• /
• 2024

The core shroud of rector vessel internals (RVI) of OPR1000 and ARP1400 is made of Type 304 stainless steel (SS) by bending and welding process that may induce high deformation and residual stress in manufacturing. This work aims to evaluate the susceptibility of stress corrosion crack (SCC) initiation of bent parts of RVI in high temperature primary water environment. For SCC initiation test, tensile specimens were fabricated from the 90 degree bent plate of Type 304 SS (DT specimen), that is an archived part of a Korean APR1400. After the SCC initiation test, the specimen surface was thoroughly examined by optical and scanning electron microscopy, and compared to the specimen fabricated from the as-received plate of Type 304 SS (AR specimen). The surface observation revealed that SCC initiated on the AR specimen surface in typical intergranular (IG) mode, while SCC on the DT specimen occurred in transgrannular mode as well as IG mode. It was also found that the size and number of SCC on the DT specimen were larger than that on the AR specimen. This was attributable to a strain-hardening during the bending process. To compare the susceptibility of SCC initiation, total crack density (TCD) was calculated from the total crack length divided by the measured area of AR and DT specimens. TCD of DT specimen was 4.6 times higher than AR specimen in average, indicating that higher possibility of degradation of bent parts of RVI for a long-term operation.

• Nuclear Engineering and Technology
• /
• v.56 no.6
• /
• pp.2002-2010
• /
• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• Nuclear Engineering and Technology
• /
• v.41 no.5
• /
• pp.617-648
• /
• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• Nuclear industry
• /
• v.26 no.4 s.278
• /
• pp.85-103
• /
• 2006

국제원자력규제자협의회(INRA) 정식 가입/ "원자력 기술 주도국으로 부상"/ "방패장 건설 사업 착실히 추진"/ 방폐장 처분 방식 "객관적이고 투명하게"/ 인도네시아 국회의장 고리 원전 방문/ 세계 원자력 안전 전문가들 한 자리에/ 아시아 태평양 지역 17개국 원자력 협력 토의/ 국제핵융합실험로(ITER) 공동이행협정문서 최종 확정/ 한 · IAEA, 원자력방호기술 향상 방안 협의/ 울진 원전 6호기 리스크 정보 활용 정기 검사 수행/ IAEA, 국내 원전 안전 점검 협의 위해 방한/ 중국 원전 기술단 방한/ 울진 원전 총발전량 4천억kWh 돌파/ 원전 안전 관리 실태 불시 점검 결과 양호/ '핵융합연구센터' 공식 출범/ 핵융합에너지개발진흥법 제정을 위한 공청회 개최/ 제9차 한-미 안전조치 기술협력 회의 개최/ 개량핵연료 'PLUS7' 첫 상용화/ 베트남 원자력 협력 강화 워크숍 개최/ "2015년엔 세계 최우수 발전회사"/ 'APR1400'중국에 적극 홍보/ 한수원, 협력기업에 500억원 지원/ "한국 원자력 도입 경험 배우겠다"/ 방사선원 사고 대비 예방 안전 체계 구축/ "한수원 신용 등급 A+"/ 울진 6호기 '첫번째' 계획 예방 정비착수/ '2006 원자력 체험전' 개최/ 원자로 내부 이물질 감시 시스템 개발/ 월성 3호기 정비 완료, 발전 재개/ IAEA안전조치이행자문단 월성 원전 방문/ 제11회 원자력안전기술정보회의 개최/ 월성 1호기 방사능 방재 전체 훈련 실시/ (주)선바이오텍, 첫 '연구소 기업' 공식 승인/ 방사선원 안전성 향상 워크숍 개최/ 양성자 가속기 사업 본격 궤도에/ 스웨덴 프로스마르크 원전 관계자 방한/ "방폐장 유치한 성숙한 시민 의식 감탄" / "올해는 혁신을 체질화하겠다"/ KAERI, 연구윤리위원회 발족/ 2005년 지식 활동 우수 그룹 포상/

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.36 no.10
• /
• pp.1043-1047
• /
• 2012

The European nuclear design requirements, which should be satisfied by nuclear reactors in Europe, usually recommend a so-called core catcher, which is a molten core ex-vessel cooling facility, to manage a severe accident at a nuclear reactor. Two different types of core catcher concepts are compared to determine their abilities to manage severe accidents and cool core melts. The study reveals that direct cooling is better for cooling capacity and is convenient to construct, while indirect cooing is better for the management of a severe accident.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.1
• /
• pp.115-125
• /
• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Journal of the Korean Society of Systems Engineering
• /
• v.12 no.1
• /
• pp.89-103
• /
• 2016

Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.

• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.1045-1052
• /
• 2009

A calculation model is developed to predict the transient free surface flow on the containment floor following a loss-of-coolant accident (LOCA) of pressurized water reactors (PWR) for the use of debris transport evaluation. The model solves the two-dimensional Shallow Water Equation (SWE) using a finite volume method (FVM) with unstructured triangular meshes. The numerical scheme is based on a fully explicit predictor-corrector method to achieve a fast-running capability and numerical accuracy. The Harten-Lax-van Leer (HLL) scheme is used to reserve a shock-capturing capability in determining the convective flux term at the cell interface where the dry-to-wet changing proceeds. An experiment simulating a sudden break of a water reservoir with L-shape open channel is calculated for validation of the present model. It is shown that the present model agrees well with the experiment data, thus it can be justified for the free surface flow with accuracy. From the calculation of flow field over the simplified containment floor of APR1400, the important phenomena of free surface flow including propagations and interactions of waves generated by local water level distribution and reflection with a solid wall are found and the transient flow rates entering the Holdup Volume Tank (HVT) are obtained within a practical computational resource.