• Nuclear Engineering and Technology
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• 제38권4호
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• pp.343-352
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• 2006

Convenient analytical tools for evaluation of the aperiodic and the fluctuating instabilities of the passive residual heat removal system (PRHRS) of an integral reactor are developed and results are discussed from the viewpoint of the system design. First, a static model for the aperiodic instability using the system hydraulic loss relation and the downcomer feedwater heating equations is developed. The calculated hydraulic relation between the pressure drop and the feedwater flow rate shows that several static states can exist with various numbers of water-mode feedwater module pipes. It is shown that the most probable state can exist by basic physical reasoning, that there is no flow rate through the steam-mode feedwater module pipes. Second, a dynamic model for the fluctuating instability due to steam generation retardation in the steam generator and the dynamic interaction of two compressible volumes, that is, the steam volume of the main steam pipe lines and the gas volume of the compensating tank is formulated and the D-decomposition method is applied after linearization of the governing equations. The results show that the PRHRS becomes stabilized with a smaller volume compensating tank, a larger volume steam space and higher hydraulic resistance of the path $a_{ct}$. Increasing the operating steam pressure has a stabilizing effect. The analytical model and the results obtained from this study will be utilized for PRHRS performance improvement.

• 디지털산업정보학회논문지
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• 제17권1호
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• pp.35-44
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• 2021

With the development of information technology, the cybersecurity threat continues as digital-related technologies are applied to the instrumentation and control system of nuclear power plants. The malfunction of the instrumentation and control system can cause economic damage due to shutdown, and furthermore, it can lead to national disasters such as radioactive emissions, so countering cybersecurity threats is an important issue. In general, the study of cybersecurity in instrumentation and control systems is concentrated on safety systems, and diverse protection systems perform protection and reactor shutdown functions, leading to reactor shutdown or, in the worst case, non-stop situations. To accurately analyze cyber threats in the diverse protection system, its linked facilities should be analyzed together. Risk analysis should be conducted by analyzing the potential impact of inter-facility cyberattacks on related facilities and the impact of cybersecurity on each configuration module of the diverse protection system. In this paper, we analyze the linkage of the diverse protection system and discuss the cybersecurity linkage threat by analyzing the availability of equipment, the cyber threat impact of the linked equipment, and the configuration module's cybersecurity vulnerability.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2670-2689
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• 2022

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1998년도 춘계 학술발표회 논문집
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• pp.235-242
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• 1998

Aiming at one of decisive alternatives for long term aspect of nuclear power concerns, an integral and closed nuclear system, AMBIDEXTER (Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor) concept is under development. The AMBIDEXTER complex essentially comprises two mutually independent loops of the radiation/material transport and the heat/energy conversion, centered at the integrated reactor assembly, which enables one to utilize maximum benefits of nuclear energy under minimum risks of nuclear radiation. And it provides precious radioisotopes and radiation sources from its waste stream. Also the reactor operates at very low level of fission products inventory throughout its lifetime. The nuclear and thermalhydraulic characteristics of the molten TH/$^{233}$ U fuel salt extend the capability of the self-sustaining AMBIDEXTER fuel cycle to enhance resource security and safeguard transparency. The reactor system is consisted of a single component module of the core, heat exchangers and recirculation pumps with neither pipe connections nor active valves in between, which will significantly improve inherent features of nuclear safety. States of the core technologies associated with designing and developing the AMBIDEXTER concept are mostly available in commercialized form and thus demonstration of integral aspects of the concept should be the prime area in future R＆D programs.

• 대한기계학회논문집A
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• 제24권8호
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• pp.2083-2090
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• 2000

In order to manage nuclear power plants safely and cost effectively, it is necessary to develop integrity evaluation methodologies for the main components. Recently, the integrity evaluation techniques were broadly studied regarding the license renewal of nuclear power plants which were approaching their design lives. Since the integrity evaluation process requires special knowledges and complicated calculation procedures, it has been allowed only to experts in the specified area. In this paper, an integrity evaluation system for reactor pressure vessel was developed. RVIES(Reactor Vessel Integrity Evaluation System) provides four specific integrity evaluation procedures covering PTS(Pressurized Thermal Shock) analysis, P-T(Pressure-Temperature) limit curve generation, USE(Upper Shelf Energy) analysis and Fatigue analysis. Each module was verified by comparing with published results.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1306-1313
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• 2018

The randomness and incipient nature of certain faults in reactor systems warrant a robust and dynamic detection mechanism. Existing models and methods for fault diagnosis using different mathematical/statistical inferences lack incipient and novel faults detection capability. To this end, we propose a fault diagnosis method that utilizes the flexibility of data-driven Support Vector Machine (SVM) for component-level fault diagnosis. The technique integrates separately-built, separately-trained, specialized SVM modules capable of component-level fault diagnosis into a coherent intelligent system, with each SVM module monitoring sub-units of the reactor coolant system. To evaluate the model, marginal faults selected from the failure mode and effect analysis (FMEA) are simulated in the steam generator and pressure boundary of the Chinese CNP300 PWR (Qinshan I NPP) reactor coolant system, using a best-estimate thermal-hydraulic code, RELAP5/SCDAP Mod4.0. Multiclass SVM model is trained with component level parameters that represent the steady state and selected faults in the components. For optimization purposes, we considered and compared the performances of different multiclass models in MATLAB, using different coding matrices, as well as different kernel functions on the representative data derived from the simulation of Qinshan I NPP. An optimum predictive model - the Error Correcting Output Code (ECOC) with TenaryComplete coding matrix - was obtained from experiments, and utilized to diagnose the incipient faults. Some of the important diagnostic results and heuristic model evaluation methods are presented in this paper.

• 한국물환경학회지
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• 제27권1호
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• pp.44-53
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• 2011

Membrane bio-reactor (MBR) has several advantages over the conventional activated sludge process, including a high biomass, low sludge production, and better permeate quality. Therefore, the MBR have gained popularity for municipal and industrial wastewater treatment. However the MBR usually were used for sewage and low streng th wastewater treatment because of membrane fouling problem and limitation of oxygen transfer into biomass. In this study, the hybrid process combining MBR and pure oxygen was tested for high strength organic wastewater treatment in the COD loading range from 2 to $10kgCOD/m^3{\cdot}day$. The hybrid process, membrane coupled pure oxygen high compact reactor (MPHCR), had been operated for one year and operation parameters, the effect of COD loading, MLSS concentration and the location of membrane module were studied for membrane fouling characteristic. Also membrane resistance test and the component of foulant was analyzed to investigate what is specific foulant in the MBR.

• 상하수도학회지
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• 제25권4호
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• pp.565-572
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• 2011

The effects of chemical pretreatments on the excess sludge production in the membrane-coupled bioreactor were investigated. In addition, their effects on membrane fouling were also evaluated. Two membrane bioreactors were operated. In one reactor, a part of the mixed liquor was t reated with NaOH and ozone gas consecutively and was returned to the reactor. T he f lowrate of the chemical pretreatment stream was 1.5% of the influent flowrate. During the 200days of operation, the MLSS level in the bioreactor with mixed liquor pretreatment was maintained relatively constant at the range of 8,000 ~ 10,000$mg/{\ell}$ while it increased steadily up to 26,000 $mg/{\ell}$ in the absence of the pretreatment. Each reactor was equipped with two laboratory membrane modules where the flux for each module was 20, and 30 ${\ell}/m^2{\cdot}h$, respectively. With pretreatment, almost constant transmembrane pressure(TMP) was observed throughout the operation at the flux of 20 ${\ell}/m^2{\cdot}h$. Without pretreatment the membrane module at the same flux could also be operated at relatively stable condition. However, as the MLSS increases up to 25,000 $mg/{\ell}$, a fast TMP increase was observed. In conclusion, a complete control of excess sludge production in the membrane-coupled bioreactor was possible without significant deterioration of the treated water quality. In addition, it was shown that stable operation in terms of TMP is possible with sludge pretreatment and recirculation.

• 한국초전도ㆍ저온공학회논문지
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• 제13권1호
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• pp.41-45
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• 2011

Hybrid fault current limiters (FCL) have been researched at Yonsei University. The hybrid FCL has advantages such as having a rapid response to a sudden fault situation and a fast recovery time from a quench. It consists of an asymmetric HTS coil, a switching module, and a bypass reactor. The asymmetric HTS coil is wound with two different types of HTS wires in an opposite direction so that it has nearly zero inductance at the superconducting state. When the quench occurs at the fault state, a strong magnetic field is generated from the asymmetric coil because of different quench characteristics of two HTS wires, and then a repulsive force is induced in the switching module. The force opens the switch and the fault current is pushed into the bypass reactor. In this research, we analyzed the cause of the repulsive force and confirmed, experimentally and computationally, that the magnitude of a repulsive force is varied by changing the gap distance between the asymmetric coil and the switching module. By using the FEM simulation, we calculated the repulsive force with respect to the gap distance and verified that the effect of the gap distance. Then, short circuit test was carried out to confirm the correct operation of the fast switch.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.47-55
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• 2005

가상의 연구로 해체 시설 환경을 설계하는데 필요한 단위 프로그램들의 특성을 검토한 후 결과 자료를 바탕으로 해체 디지털 목업 시스템의 설계가 완료되었다. 단위 프로그램들은 해체 데이터베이스 시스템 모듈 연구로 시설과 제염 및 해체 장비를 3차원으로 모델링하는 모듈, 3차원으로 방사능 오염 분포도를 묘사하는 모듈, 그리고 해체 일정 및 해체 비용을 평가하는 모듈 등으로 구분된다. 독립적으로 운영되는 이들 단위 모듈들을 통합된 시스템으로 만들기 위해 단위 모듈들의 아키텍쳐 설계 연구가 수행되었다. 연구 결과 다양한 모듈들로 구성된 해체 디지털 목업 시스템을 통합된 환경에서 시나리오를 시험 평가할 수 있도록 하기위해 연구로 시설의 제염 및 해체 활동을 시각적으로 보여줄 수 있는 가시화(visualization) 모듈과 해체 일정 및 해체 비용을 평가하고 분석하는 시뮬레이션(simulation) 모듈로 해체 디지털 목업 시스템의 아키텍쳐를 구현하였다.