• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• 대한기계학회논문집A
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• 제35권10호
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• pp.1265-1271
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• 2011

본 논문에서는 유한요소 모델을 사용한 유체-구조 해석을 통하여 원자력 발전소 임시 받침대의 내진에 대한 안전성을 평가하였다. 임시받침대는 수중에 존재하기 때문에 유체-구조 연성을 통하여 유체의 영향을 고려하였다. 유체의 영향은 구조물의 단위길이당 추가질량으로 정의하여 적용하였다. 각각의 운전기준지진(OBE)과 안전정지지진(SSE)의 설계조건을 층응답스펙트럼(Floor Response Spectrum: FRS)으로 적용하여 진동해석과 내진해석을 수행하였다. 해석된 임시받침대의 최대변위는 운전기준지진에서 0.29mm 이고, 운전정지지진에서 최대변위는 0.36 mm 이다. 최대응력은 운전 기준지진에서 17.9 MPa, 안전정지지진에서 19.6 MPa 이며, 이 값은 재료의 항복강도의 23 %, 14 % 수준이다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• 한국정보통신학회:학술대회논문집
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• 한국해양정보통신학회 2011년도 추계학술대회
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• pp.411-414
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• 2011

본 논문에서는 DPSS 기능을 갖는 3중 모드 DC-DC buck 변환기를 설계하였다. 설계된 buck 변환기는 부하 전류가 큰 경우(80mA~500mA)에는 PWM(Pulse Width Modulation) 제어 방식을 사용하고, 부하 전류가 작은 경우(1mA~80mA)에는 PFM(Pulse Frequency Modulation) 제어 방식을 사용하며, 부하 전류가 1mA 이하인 대기모드(sleep-mode)에서는 LDO(Low Drop Out)를 사용한다. 또한, PFM 제어 방식에서 부하 전류가 작은 경우 효율을 증가시키기 위해 DPSS(Dynamic Partial Shutdown Strategy) 기법을 사용하였다. 그 결과 넓은 부하 전류 범위에서 높은 효율을 얻을 수 있다. 제안된 벅 변환기는 CMOS 0.18um 공정을 이용하여 설계되었다. 3.3V의 입력전압을 받아 2.5V의 출력전압으로 강압시키며, 최대 부하전류는 500mA이고, 스위칭 주파수는 1MHz이다. 최대효율은 97.03 %, 칩 크기는 PAD를 포함하여 $1465um{\times}895um$이다.

• 한국정보과학회논문지:소프트웨어및응용
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• 제27권2호
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• pp.120-133
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• 2000

산업현장에서 복합 실시간 계통(HRTS: Hybrid Real-Time Systems) 개발을 위한 정형기법 사용의 주된 장벽은 인지적 어려움이며 이는 또 다른 위험을 초래할 수 있다. 이러한 문제를 극복하기 위해 HRTS 요구분석과 안전성 분석 시 사용자의 인지적 부담을 줄여줄 수 있는 정성적 요구분석 체계를 제안한다. 이 체계는 요구사항 명세를 위한 정성적 정형기법(QFM: Qualitative Formal Method)과 인과정보에 의한 요구사항 안전성 분석기법(CRSA: Causal Requirements Safety Analysis)으로 구성되어 있다. QFM에서는 인공지능 분야에서 연구된 정성추론 이론을 정형명세에 도입하여 요구사항 설계자와 분석자의 인지적 부담을 줄일 수 있도록 하였다. CRSA는 QFM에서 도출한 HRTS 동작의 인과 정보에 따라 체계적으로 위험 원인을 추적할 수 있도록 하여, 기존 결함 트리 분석(FTA: Fault Tree Analysis) 기법의 단점인 분석자의 주관에 의존하는 문제를 해결한다. 월성 원자력 발전소 자동정지계통(Shutdown System 2) 소프트웨어 요구사항 명세와 안전성 분석에 QFM과 CRSA를 적용하여 그 실효성을 입증하고자 하였다.

• 한국산학기술학회논문지
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• 제16권9호
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• pp.6297-6304
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• 2015

우리나라 최초의 원자력발전소인 고리 원전 1호기의 폐쇄가 결정됨에 따라 원전 해체가 화두가 되고 있다. 원전 해체는 우리나라에서 한번도 경험해 보지 못한 일로 해체 과정도 어렵고 시간도 많이 소요된다. 그 일부분인 해체물량 또한 파악이 어렵다. 본 연구에서는 최근 건설산업에 많이 활용되고 있는 BIM 기술을 원전 해체 물량산출에 활용할 수 있는 방안을 제시하였다. 그 결과, 원전 해체 공법선정 및 공정 확립, BIM 모델링 환경 준비, 작업분류체계 구축, 객체분류체계 구축, BIM 통합모델 작성, BIM 객체에 물량 속성 배분 등의 방안을 제시하였다. 제시한 방안은 영구정지 대상 원전이 집중적으로 발생하는 시기부터 유용하게 활용될 수 있다. 이에 기반한 기술확보를 통해 나아가 해외 원전 해체 사업 수주에도 유리하게 작용할 것으로 기대된다.

• 한국화재소방학회논문지
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• 제24권4호
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• pp.116-122
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• 2010

원자력발전소는 심층화재방어 개념에 따라 화재 발생시 발전소 외부로 방사능의 누출을 억제하고, 발전소의 안전정지 기능이 유지될 수 있도록 설계, 건설, 운영되어야 한다. 해외의 원전건설 국가는 이러한 원자력발전소의 안전정지 기능의 목적을 달성하기 위하여 원자력발전소에 대한 통합된 화재방호 규제요건을 가지고 있으나, 우리나라의 경우, 원자력발전소의 화재방호계통을 적용하기 위한 강제 요건으로서 소방관계법과 원자력법을 동시에 적용하는 비합리적인 규제지침을 가지고 있다. 화재방호설비에 관하여 원자력발전소 운영에 오랜 경험을 가진 미국, 캐나다 및 일본의 기술적으로 단일화된 원자력발전소의 화재방호 규제체계를 제시하였고, 우리나라도 화재방호설비는 원자력법에 의한 화재하중에 따른 화재위험도분석 결과를 설계에 반영하여 소방관계법에서는 예외조항으로 인정받을 수 있어야 함을 제안하였다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.575-600
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• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• 한국지진공학회논문집
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• 제24권1호
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• pp.1-8
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• 2020

In 2016, an earthquake occurred at Gyeongju, Korea. At the Wolsong site, the observed peak ground acceleration was lower than the operating basis earthquake (OBE) level of Wolsong nuclear power plant. However, the measured spectral acceleration value exceeded the spectral acceleration of the operating-basis earthquake (OBE) level in some sections of the response spectrum, resulting in a manual shutdown of the nuclear power plant. Analysis of the response spectra shape of the Gyeongju earthquake motion showed that the high-frequency components are stronger than the response spectra shape used in nuclear power plant design. Therefore, the seismic performance evaluation of structures and equipment of nuclear power plants should be made to reflect the characteristics of site-specific earthquakes. In general, the floor response spectrum shape at the installation site or the generalized response spectrum shape is used for the seismic performance evaluation of structures and equipment. In this study, a generalized response spectrum shape is proposed for seismic performance evaluation of structures and equipment for nuclear power plants. The proposed response spectrum shape reflects the characteristics of earthquake motion in Korea through earthquake hazard analysis, and it can be applied to structures and equipment at various locations.