• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 시스템엔지니어링학술지
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• 제14권2호
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• pp.24-32
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• 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.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.389-401
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• 2024

In recent decades, numerous instances of blockages have been reported in coastal nuclear power plants globally, leading to serious safety accidents such as power reduction, manual or automatic power loss, or shutdown of nuclear power units. Loss or shortage of cooling water may compromise the reliability of the cooling water system, thus threatening the operational safety of power plants and resulting in revenue reduction. This study provides a comprehensive review of the current state of cooling water system safety in coastal nuclear power plants worldwide and the common challenges they face, as well as the relevant research on cooling water system safety issues. The research overview and progress in investigation methods, outbreak mechanisms, prevention and control measures, and practical cases of blockages were summarized. Despite existing research, there are still many shortcomings regarding the pertinence, comprehensiveness and prospects of related research, and many problems urgently need to be solved. The most fundamental concern involves understanding the list of potential risks of blockages and their spatially distributed effects in surrounding waters. Furthermore, knowledge of the biological cycles and ecological habits of key organisms is essential for implementing risk prevention and control and for building a scientific and effective monitoring system.

• 한국정보과학회논문지:컴퓨팅의 실제 및 레터
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• 제8권1호
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• pp.46-61
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• 2002

소프트웨어의 개발 단계 중 품질을 결정하는 주요 단계는 요구 명세 단계로 알려져 있다. 따라서, 소프트웨어 개발 업체는 소프트웨어 요구명세서의 분석을 가장 중요한 단계 중 하나로 취급하고 있고, 특히 안전성이 중요한 시스템의 경우에는 시스템을 운영하기 위하여 국내와 국제적인 규제 기관에서는 요구 명세의 분석을 통한 안전성의 입증을 요구한다. 소프트웨어의 요구 명세 분석을 위한 방법 중 인스펙션과 정형 검증이 가장 효과적인 방법으로 알려져 있다. 본 논문에서는 SCR-style의 요구 명세를 정리 증명기인 PVS를 이용하여 정형 검증을 수행하는 방법을 제안하였다. 그리고, 논문에서 제안된 방법으로 실제 월성 원자력 발전소의 정지 시스템의 검증을 수행하였다. 이 시스템은 인스펙션으로 검증된 적은 있으나 정형 검증 방법으로는 증명된 적이 없고, 국내에서 실제 운영되는 산업계시스템에 정형 검증 방법이 적용된 사례는 매우 드물기 때문에 차후 정형 검증 방법을 적용하기 위한 평가로서도 이와 같은 실험적인 적용이 매우 중요하다.

• 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.

• 한국가스학회지
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• 제24권3호
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• pp.54-62
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• 2020

구미 휴브글로벌 AHF 누출사고를 보면서 화학 사고가 우리 사회에 미치는 영향이 얼마나 큰지 경험한 바 있다. 이후로 여러 분야에서 화학 사고에 관한 많은 연구가 수행되고 있다. 사용처에서는 공정 시스템에 대한 개선방안을 찾아 현장에 적용하고자 했고 안전분야에서는 피해 영향 분석을 통한 안전성 향상을 연구하여 비상대응에 적용하여 피해 영향을 줄이고자 하였다. 본 연구에서는 이충전 작업 시 일어나는 화학사고에 신속하게 대응할 수 있는 기계적 안전장치를 적용하여 AHF ISO Tank의 안전성을 높이고자 하였다. 연구의 수행으로는 우리나라에 연간 유통되는 AHF의 수입량의 규모와 운반의 수단으로 사용되는 ISO Tank의 운송 물량을 조사하고 AHF ISO Tank의 구조 및 Unloading 절차를 조사하였다. 많은 유형의 탱크로리를 조사하는 과정에서 AHF ISO Tank 구조와 작업 절차가 유사한 염소 탱크로리에 기계적 안전장치인 EFV(과류 방지 밸브)가 설치되어 있는 것을 확인하였다. EFV의 AHF ISO Tank에 적용성과 도입 시 안전기능인 긴급차단에 대한 성능을 휴브글로벌 사고와 2018년에 울산에서 발생한 염소 Tank lorry의 사고사례를 비교 조사하여 실제 사고에서 얻은 자료로 검정할 수 있었다. 사고사례와 적용성 등을 비교한 결과, AHF ISO Tank에 화학사고 피해를 감소시키고 사고를 조기에 수습할 할 수 있는 기계적 안전장치인 EFV를 도입할 것을 제안하며 휴브글로벌 사고와 같은 화학 사고에 도움이 되고자 한다.

• Journal of international Conference on Electrical Machines and Systems
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• 제2권4호
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• pp.460-465
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• 2013

Electrical equipments of nuclear power plant are divided into class 1E and non-class 1E. Electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, are classified as class 1E. batteries of nuclear power plant are divided into four channels, which are physically and electrically separate and independent. The battery bank of class 1E DC power system of the nuclear power plant use lead-acid batteries in present. The lead acid battery, which has a high energy density, is the most popular form of energy storage. Kt factor of lead-acid battery is used to determine battery size and it is one of calculatiing coefficient for capacity. this paper analyzes Kt factor of lead-acid battery for the DC power system of nuclear power plant. In addition, correlation between Kt parameter and peukert's exponent of lead-acid battery for nuclear plant are discussed. The analytical results contribute to optimize of determining size Lead-acid battery bank.

• 터널과지하공간
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• 제10권2호
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• pp.249-256
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• 2000

1980년대 말 이후 산업구조의 개편에 따라 국내 대부분의 광산은 폐·휴업 상태이며, 기존 채굴공동은 방치되고 있어 부분적으로 환경 저해요인으로 인식되고 있는 실정이다. 따라서,이들 폐공동의 활용방안을 검토하기 위하여 1차적으로 외국의 활용 사례를 검토하였다. 그 결과 농·수·축산물의 지하저장, 압축공기의 저장시설, 산업 폐기물의 처리장으로의 활용 예를 예시하였다. 그러나 국내에서 비교적 쉽게 활용할 수 있는 분야는 산업 폐기물의 처리장이라 할 수 있다. 이를 위해서는 채굴공동에 대한 보강 및 안전 대책의 수립은 물론 각 이용 목적에 따른 암반공학적 연구가 선행되어야 한다.

• 한국전산유체공학회지
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• 제17권4호
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• pp.1-8
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• 2012

During a reactor normal operation, two parallel 50% capacity cooling pumps circulate primary coolant to remove the fission reaction heat of the reactor through heat exchangers cold by a cooling tower. When one pump is failure, the other pump shall continuously circulate the coolant to remove the residual heat generated by the fuels loaded in the reactor after reactor shutdown. It is necessary to estimate how much flow rate will be supplied to remove the residual heat. We carried out a flow network analysis for the parallel primary pumps based on the piping network of the primary cooling system in HANARO. As result, it is estimated that the flow rate of one pump increased about 1.33 times the rated flow of one pump and was maintained within the limit of the cavitation critical flow.