• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• 한국지반공학회논문집
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• 제33권7호
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• pp.55-64
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• 2017

심층 처분방식은 고준위폐기물을 처분하기 위한 가장 적합한 대안으로 고려되어지고 있다. 심층 처분시설은 지하 500~1,000m 깊이의 암반층에 설치되며 심층 처분시스템의 구성 요소로는 처분용기, 완충재, 뒷채움 및 근계 암반이 있다. 이 중 완충재는 심층 처분시스템에 있어 매우 중요한 역할을 한다. 완충재는 지하수 유입으로부터 처분용기를 보호하고, 방사성 핵종 유출을 저지한다. 처분용기에서 발생하는 고온의 열량이 완충재로 전파되기에 완충재의 열적 성능은 처분시스템의 안정성 평가에 매우 중요하다고 할 수 있다. 따라서 본 연구에서는 국내 경주산 압축 벤토나이트 완충재에 대한 열전도도 추정 모델을 개발하고자 하였다. 압축 벤토나이트 완충재의 열전도도는 비정상 열선법을 이용하여 다양한 함수비와 건조밀도에 따라 측정하였으며, 총 39개의 실험 데이터를 토대로 회귀분석을 이용하여 경주 압축 벤토나이트의 열전도도 추정 모델을 제시하였다.

• 한국산학기술학회논문지
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• 제12권2호
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• pp.821-828
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• 2011

본 논문은 자동항로 추적(Track keeping control), 자동조타(Automatic steering), 자동 접이안(Automatic mooring control) 등으로 구성된 자동운항 시스템 중 자동조타장치의 성능 개선 알고리즘 개발에 대해 다루고 있다. 자동조타는 풍력 또는 조력 등의 영향으로부터 선박의 설정 항로와 실제 침로와의 차이를 계산하여 설정된 항로를 유지하며 항해하므로, 조타에 소요되는 선원의 지속적인 항해로 인한 운전 부담을 경감시키고 불필요한 타조작에 의한 항로 이탈을 줄임으로써 항해거리 단축과 연료비용을 절약할 수 있는 시스템이다. 선박의 모델링을 위하여 Nomoto 모델에 근거하여 전달함수를 구하고, 조종성능(Manoeuvirng) 편리성을 고려하여 타각 입력에 대한 선수각 응답으로 표시된 선박의 4자유도만을 고려한 선형 모델을 제안하고 선박 자동조타장치의 최대각과 타각율을 고려하여 Fuzzy제어기를 설계 하였고 PID제어기로 성능을 비교 분석하였다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• 청정기술
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• 제25권1호
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• pp.19-32
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• 2019

여러 국가를 통행하는 다양한 형태의 선박에서 배출되는 황산화물에 의해 지역적인 대기오염이 심화됨에 따라 국제해사기구에서는 황산화물 배출제어지역을 설정하여 규제하고 있다. 이러한 지역적 규제를 만족시키기 위해서는 선박의 연료선택방법과 배가스 후처리 장치가 적용되고 있으나 경제적인 이유로 스크러버를 설치하여 배출되는 황산화물의 양을 저감하는 배가스 후처리 방법이 주로 선호되고 있다. 스크러버는 배출가스 중 황산화물을 액적에 흡수시켜 황산화물의 양을 저감하는 장치로 액적의 크기에 따라 스크러버의 성능이 좌우된다. 이러한 성능을 평가하기 위해서, 본 논문에서는 대향류형 스크러버와 사이클론 스크러버를 설계하고, 전산유체역학을 이용하여 각 액적의 크기에 따른 탈황 효율과 액적이 증발되는 양을 평가하였다. 평가 방법으로 스크러버 내부는 기체와 액체의 2상 유동을 가지기 때문에, Eulerian-Eulerian 해석 기법을 사용하였으며, 액적의 직경이 $100{\mu}m$, $300{\mu}m$, $500{\mu}m$와 $700{\mu}m$일 때 계산을 진행하여 스크러버를 분석하였다. 계산 결과, 2종류의 스크러버 모두 $500{\mu}m$와 $700{\mu}m$일 때 높은 탈황 효율과 낮은 증발량을 나타내었다.

• 한국분말재료학회지
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• 제29권5호
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• pp.411-417
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• 2022

Environmental issues such as global warming due to fossil fuel use are now major worldwide concerns, and interest in renewable and clean energy is growing. Of the various types of renewable energy, green hydrogen energy has recently attracted attention because of its eco-friendly and high-energy density. Electrochemical water splitting is considered a pollution-free means of producing clean hydrogen and oxygen and in large quantities. The development of non-noble electrocatalysts with low cost and high performance in water splitting has also attracted considerable attention. In this study, we successfully synthesized a NiCo₂O₄/NF electrode for an oxygen evolution reaction in alkaline water splitting using a hydrothermal method, which was followed by post-heat treatment. The effects of heat treatment on the electrochemical performance of the electrodes were evaluated under different heat-treatment conditions. The optimized NCO/NF-300 electrode showed an overpotential of 416 mV at a high current density of 50 mA/cm² and a low Tafel slope (49.06 mV dec^-1). It also showed excellent stability (due to the large surface area) and the lowest charge transfer resistance (12.59 Ω). The results suggested that our noble-metal free electrodes have great potential for use in developing alkaline electrolysis systems.

• Journal of Advanced Marine Engineering and Technology
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• 제40권3호
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• pp.157-164
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• 2016

본 연구에서는 SI 공정 시스템에서 헬륨 가열 장치에 대한 설계와 건조 데이터 확보를 위한 파이롯 규모의 헬륨 가열 장치 시스템에 대한 예비 설계 및 해석을 수행하였다. 헬륨 가열기는 LPG연소로를 활용하도록 설계하였고, 열유동 해석을 수행한 결과, 열교환기 출구측에서 LPG 연소가스의 유속이 약 40 m/s가 되었다. 최대온도는 6개의 베플이 설치된 경우가 4개의 베플이 설치된 경우보다 높게 나타나며, 이는 6개의 베플일 때 연소가스에서 헬륨가스로의 열전달이 좋아질 것임을 의미한다. 더불어, 베플수가 많아지면 LPG 연소가스의 유량을 감소시킬 수 있어 연료비용을 저감할 수도 있음을 의미한다. 다만, 베플수를 무한정 증가시키면 입출구의 압력차가 증가되기 때문에 최적의 베플수 선정이 중요하다. 열유동 해석에 이어 수행한 구조해석에서는 헬륨의 유량을 3.729 mol/s, 출구 온도를 $910^{\circ}C$로 유지할 경우 관의 양 끝단에서 지지하는 경우 중간부분의 팽창률이 3.86 mm임을 확인하였다. LPG 연소로 헬륨 가열시스템에서 shell & tube type의 열교환기를 적용하여 $135^{\circ}C$의 헬륨이 $910^{\circ}C$로 가열하여 유출하기 위해서 약 $1300^{\circ}C$의 연소가스 온도 및 52 g/s의 연소가스 유량이 확보되어야 함을 확인하였다.

• 방사성폐기물학회지
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• 제9권4호
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• pp.207-217
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• 2011

UV-Vis 분광광도법과 시간분해 레이저 유도 형광분광법(TRLFS)을 이용하여 흄산의 모사 리간드로 사용한 2,6-Dihydroxybenzoate(DHB)와 U(VI)의 착물형성반응을 조사하였다. U(VI)-DHB 착물 고유의 전하이동 흡수 스펙트럼을 분석한 결과, 착물형성반응은 우라늄-리간드 비가 1:1 또는 1:2 착물을 형성하는 이중 평형반응이며, 산도에 따라 착물종의 분포가 변한다는 것을 밝혔다. 계산된 착물형성상수 (log $K_1$ and log $K_2$)는 $12.4{\pm}0.1$과 $11.4{\pm}0.1$이다. 이에 더하여, TRLFS 방법으로 조사한 결과, DHB는 U(VI) 화학종들의 형광 소광제(quencher)로서 역할을 한다는 것을 확인하였다. 특히, 확인된 U(VI) 화학종 모두(${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$과 $(UO_2)_3(OH)_5{^+})$에서 정적 (static) 및 동적 (dynamic) 소광작용이 공존하는 것으로 관찰되었다. 시간분해 형광 스펙트럼으로부터 리간드 농도에 따른 U(VI) 화학종의 형광세기와 형광수명을 측정하였으며, Stern-Volmer 식을 이용하여 분석하였다. 결정된 정적소광계수(KS)는 ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$ 과 $(UO_2)_3(OH)_5+$에 대하여 각각 $4.2{\pm}0.1$, $4.3{\pm}0.1$ 과 $4.34{\pm}0.08$이다. Stern-Volmer 식을 이용한 분석 결과, 단일 또는 이중 배위자 구조(mono- and bi-dentate)의 U(VI)-DHB 착물이 모두 정적소광효과에 관여하는 바닥상태 착물임을 확인하였다.

• 한국지능시스템학회:학술대회논문집
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• 한국퍼지및지능시스템학회 1993년도 Fifth International Fuzzy Systems Association World Congress 93
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• pp.1051-1054
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• 1993

The International Atomic Energy Agency's Statute in Article III.A.5 allows it“to establish and administer safeguards designed to ensure that special fissionable and other materials, services, equipment, facilities and information made available by the Agency or at its request or under its supervision or control are not used in such a way as to further any military purpose; and to apply safeguards, at the request of the parties, to any bilateral or multilateral arrangement, or at the request of a State, to any of that State's activities in the field of atomic energy”. Safeguards are essentially a technical means of verifying the fulfilment of political obligations undertaken by States and given a legal force in international agreements relating to the peaceful uses of nuclear energy. The main political objectives are: to assure the international community that States are complying with their non-proliferation and other peaceful undertakings; and to deter (a) the diversion of afeguarded nuclear materials to the production of nuclear explosives or for military purposes and (b) the misuse of safeguarded facilities with the aim of producing unsafeguarded nuclear material. It is clear that no international safeguards system can physically prevent diversion. The IAEA safeguards system is basically a verification measure designed to provide assurance in those cases in which diversion has not occurred. Verification is accomplished by two basic means: material accountancy and containment and surveillance measures. Nuclear material accountancy is the fundamental IAEA safeguards mechanism, while containment and surveillance serve as important complementary measures. Material accountancy refers to a collection of measurements and other determinations which enable the State and the Agency to maintain a current picture of the location and movement of nuclear material into and out of material balance areas, i. e. areas where all material entering or leaving is measurab e. A containment measure is one that is designed by taking advantage of structural characteristics, such as containers, tanks or pipes, etc. To establish the physical integrity of an area or item by preventing the undetected movement of nuclear material or equipment. Such measures involve the application of tamper-indicating or surveillance devices. Surveillance refers to both human and instrumental observation aimed at indicating the movement of nuclear material. The verification process consists of three over-lapping elements: (a) Provision by the State of information such as - design information describing nuclear installations; - accounting reports listing nuclear material inventories, receipts and shipments; - documents amplifying and clarifying reports, as applicable; - notification of international transfers of nuclear material. (b) Collection by the IAEA of information through inspection activities such as - verification of design information - examination of records and repo ts - measurement of nuclear material - examination of containment and surveillance measures - follow-up activities in case of unusual findings. (c) Evaluation of the information provided by the State and of that collected by inspectors to determine the completeness, accuracy and validity of the information provided by the State and to resolve any anomalies and discrepancies. To design an effective verification system, one must identify possible ways and means by which nuclear material could be diverted from peaceful uses, including means to conceal such diversions. These theoretical ways and means, which have become known as diversion strategies, are used as one of the basic inputs for the development of safeguards procedures, equipment and instrumentation. For analysis of implementation strategy purposes, it is assumed that non-compliance cannot be excluded a priori and that consequently there is a low but non-zero probability that a diversion could be attempted in all safeguards ituations. An important element of diversion strategies is the identification of various possible diversion paths; the amount, type and location of nuclear material involved, the physical route and conversion of the material that may take place, rate of removal and concealment methods, as appropriate. With regard to the physical route and conversion of nuclear material the following main categories may be considered: - unreported removal of nuclear material from an installation or during transit - unreported introduction of nuclear material into an installation - unreported transfer of nuclear material from one material balance area to another - unreported production of nuclear material, e. g. enrichment of uranium or production of plutonium - undeclared uses of the material within the installation. With respect to the amount of nuclear material that might be diverted in a given time (the diversion rate), the continuum between the following two limiting cases is cons dered: - one significant quantity or more in a short time, often known as abrupt diversion; and - one significant quantity or more per year, for example, by accumulation of smaller amounts each time to add up to a significant quantity over a period of one year, often called protracted diversion. Concealment methods may include: - restriction of access of inspectors - falsification of records, reports and other material balance areas - replacement of nuclear material, e. g. use of dummy objects - falsification of measurements or of their evaluation - interference with IAEA installed equipment.As a result of diversion and its concealment or other actions, anomalies will occur. All reasonable diversion routes, scenarios/strategies and concealment methods have to be taken into account in designing safeguards implementation strategies so as to provide sufficient opportunities for the IAEA to observe such anomalies. The safeguards approach for each facility will make a different use of these procedures, equipment and instrumentation according to the various diversion strategies which could be applicable to that facility and according to the detection and inspection goals which are applied. Postulated pathways sets of scenarios comprise those elements of diversion strategies which might be carried out at a facility or across a State's fuel cycle with declared or undeclared activities. All such factors, however, contain a degree of fuzziness that need a human judgment to make the ultimate conclusion that all material is being used for peaceful purposes. Safeguards has been traditionally based on verification of declared material and facilities using material accountancy as a fundamental measure. The strength of material accountancy is based on the fact that it allows to detect any diversion independent of the diversion route taken. Material accountancy detects a diversion after it actually happened and thus is powerless to physically prevent it and can only deter by the risk of early detection any contemplation by State authorities to carry out a diversion. Recently the IAEA has been faced with new challenges. To deal with these, various measures are being reconsidered to strengthen the safeguards system such as enhanced assessment of the completeness of the State's initial declaration of nuclear material and installations under its jurisdiction enhanced monitoring and analysis of open information and analysis of open information that may indicate inconsistencies with the State's safeguards obligations. Precise information vital for such enhanced assessments and analyses is normally not available or, if available, difficult and expensive collection of information would be necessary. Above all, realistic appraisal of truth needs sound human judgment.