• 비파괴검사학회지
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• 제25권5호
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• pp.343-349
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• 2005

광섬유 브래그 격자(fiber Bragg grating, FBC)는 1989년 Melts의 실험이후 광통신 분야 및 센서 분야에 다양한 용도를 가지고 있다. 이를 이용하여 광섬유 레이저, 대역 통과 필터로 사용하거나, 여러 가지 물리량을 측정하기 위한 온도, 변형률, 변위센서 등에 응용할 수 있다. 특히 본 연구에서는 기존 센서가 가지지 못한 긴 게이지 길이에서의 측정을 위해서 광섬유 센서 정적 시스템(FBG Static Logger)과 FBG 센서가 포함된 긴 길이의 광섬유를 이용하여 장대 구조물의 변형이나 외부 reference의 상대변위를 길게 연결하여 측정하기 쉽지 않은 변위를 측정하여 구조물의 상황을 상시 점검하고 더 나아가 잔존수명을 예측하고 보수시기를 알려줄 수 있는 방법을 모색하여 FBG. 센서의 패키지(package)를 고안하였다. 이를 통해 효율적인 장거리 변위 측정방법을 정립하고 이를 원자력 격납구조물에 적용하여 성공적으로 측정함으로써 광섬유 격자센서를 이용하여 50m 이상의 장거리 계측의 가능성을 보여 주었다.

• 에너지공학
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• 제23권3호
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• pp.186-190
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• 2014

원자력발전소는 냉각재상실사고(LOCA)와 같은 과도상태시 pH 조절을 통해 격납건물의 핵분열생성물(요오드) 제거 능력을 유지한다. 이와 더불어 격납건물 내부의 스테인레스강 기기들의 응력부식균열(Stress Corrosion Cracking)을 방지하고 알루미늄 또는 아연 부식에 의한 수소생성을 최소화할 수 있기 때문에 살수 및 집수조냉각수의 화학조건(pH) 조절능력이 요구된다. 현재 원전은 LOCA시 능동형 살수첨가제인 NaOH를 사용하여 격납건물 살수 및 집수조냉각수의 pH를 조절하도록 설계되어있다. 본 논문에서는 LOCA시 집수조냉각수의 pH를 분석하고, 살수화학조건 pH 관련 최신규제요건인 표준심사지침(SRP) 6.5.2에 따라 핵분열생성물제거상수 및 제염계수를 계산하였다. 분석결과, 격납건물집수조 pH는 8.09~9.67로서 설계기준을 만족한다. 그리고 격납건물살수계통에 의한 핵분열생성물 제거상수 및 제염계수는 원전 내환경기기검증을 위한 방사선환경 평가의 입력으로 제공된다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.795-802
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• 2013

For the improvement of APR1400 Diverse Protection System (DPS) design, the Advanced DPS (ADPS) has recently been developed to enhance the fault tolerance capability of the system. Major fault masking features of the ADPS compared with the APR1400 DPS are the changes to the channel configuration and reactor trip actuation equipment. To minimize the fault occurrences within the ADPS, and to mitigate the consequences of common-cause failures (CCF) within the safety I&C systems, several fault avoidance design features have been applied in the ADPS. The fault avoidance design features include the changes to the system software classification, communication methods, equipment platform, MMI equipment, etc. In addition, the fault detection, location, containment, and recovery processes have been incorporated in the ADPS design. Therefore, it is expected that the ADPS can provide an enhanced fault tolerance capability against the possible faults within the system and its input/output equipment, and the CCF of safety systems.

• 에너지공학
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• 제7권1호
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• pp.122-130
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• 1998

A simplified heat transfer model for the cooling capability of the AP 600 PCCS is proposed I this paper. As the PCCS domain is covered with very thin and long water film, it is phenomenologically divided into 3 regions; water entrance effect region, asymptotic region, and air entrance effect region. As the length of the asymptotic region is estimated to be over 90% of the whole domain, the phenomena in the asymptotic region is focused. Using the analogy between heat and mass transfer phenomena in a turbulent situation, a new dependent variable combining temperature and vapor mass fraction was defined. The similarity between the PCCs phenomena in the asymptotic region and the buoyant air flow phenomena on a vertical heated plate is derived. Using the similarity, the simplified heat transfer correlations for the interfacial heat fluxes and the ratios of latent heat transfer to sensible heat transfer were established. To verify the accuracy of the correlation, the results of this study were compared with those of other numerical analyses performed for the same configuration and they are well within the range of 15% difference.

• 한국도로학회논문집
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• 제15권1호
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• pp.65-73
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• 2013

PURPOSES: This paper presents the results of computer simulations of roadside safety barrier, called by safety roller guardrail, consisting of rotational roller, rotation control plate, post and subsidiary members. The rotation roller and rotation control plate are made by EVA(ethylene vinyl acetate), and PE(polyester), respectively. METHODS: The occupant risk analysis has been carried out under vehicle crash condition for high containment level of SB-4 for the purpose of local road. Simulations are performed with the finite element code LS/DYNA-3D. RESULTS: The numerical results obtained by LS/DYNA-3D software from the viewpoints of vehicle stability, vehicle trajectory, occupant risk, etc. CONCLUSIONS: It is noted that not only impact severity is drastically reduced but also vehicle trajectory is improved due to the characteristics of energy absorption and rotation pattern of EVA rollers connected by control plates.

• 에너지공학
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• 제24권1호
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• pp.42-50
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• 2015

신형 원전의 피동격납건물냉각계통(PCCS: Passive Containment Cooling System)을 구성하는 단일 전열관의 열제거 성능을 평가하기 위해, 비응축성 기체 존재 시 수직 튜브 외벽에서 발생하는 증기의 응축 열전달에 대한 실험을 수행하였다. 외경 40 mm, 길이 1.0 m의 전열관 외벽에서 증기-공기 혼합물의 평균 열전달계수를 측정하였으며, 압력 2-4 bar, 공기의 질량분율 0.1-0.7의 범위에서 실험데이터를 수집하였다. 이를 통해 압력과 비응축성기체의 농도가 응축 열전달계수에 미치는 영향을 평가하였다. 실험결과를 기존의 열전달모델인 Uchida와 Dehbi의 상관식과 비교하였으며, 이들 상관식은 실험결과에 비해 상대적으로 열전달계수를 낮게 예측함을 확인하였다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

• 한국안전학회지
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• 제27권6호
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.617-648
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• 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.