• Nuclear Engineering and Technology
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• 제47권2호
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• pp.157-164
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• 2015

Nuclear power plants have logical loop structures in their system configuration. This paper explains the method to solve a loop structure in reliability analysis. As examples of loop structured systems, the reactor core isolation cooling system and high-pressure core injection system of a boiling water reactor are considered and analyzed under a station blackout accident condition. The analysis results show the important role of loop structures under severe accidents. For the evaluation of the safety of nuclear power plants, it is necessary to accurately evaluate a loop structure's reliability.

• 한국시스템다이내믹스연구
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• 제10권1호
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• pp.109-129
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• 2009

This paper explores the ways by which the authority concerned with nuclear energy policy-making in Korea can strengthen its organisational sustainability from long-term perspective. In doing so, it applies the system dynamics approach to predict what would happen to the organisational sustainability of the nuclear energy authority in the future. In the process of analysis, it also draws causal loop map of components contained in the simulation model and constructs user-interface simulation model. It shows different predicted future values regarding organisational sustainability of the nuclear energy authority in Korea and puts forward some policy implications for practitioners and academics involved in nuclear energy policy.

• 한국소음진동공학회논문집
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• 제26권4호
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• pp.469-474
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• 2016

Abnormal sound was heard from a heat exchanger of condensate water system in a nuclear power plant, which was identified as impact sound of a loose part later. Nuclear power plants are normally equipped with loose part monitoring system for primary water system, but not for secondary water system. The abnormal sound was analyzed by using the impact signal-processing methodology based on the Hertz theory. The predicted results for impact location and size of the loose part showed good agreement with those of the actual loose part found during the overhaul period in the plant. So, this analysis methodology for the impact signal will be widely utilized for the primary and secondary side of the nuclear power plant.

• 동력기계공학회지
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• 제17권4호
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• pp.31-35
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• 2013

The emergency AC power supply system of the nuclear power plant is designed to supply the power to the nuclear power plant at the emergency operating condition. The safety function of the diesel generator at the nuclear power plant is to supply AC electric power to the safety system whenever the preferred AC power supply is unavailable. The reliable operation of onsite standby diesel generator should be ensured by a condition monitoring system designed to maintain, monitor and forecast the reliability level of diesel generator. The purpose of this paper is to improve the existing ultrasonic sensor used for condition diagnosis of engine fuel pump and cylinder head for the accurate diagnosis in actual engine condition of emergency diesel generator(EDG). As a result of this study, we could design and develop much more reliable ultrasonic sensor than existing ones.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3879-3891
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• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

• 한국정보통신학회:학술대회논문집
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• 한국정보통신학회 2022년도 춘계학술대회
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• pp.187-188
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• 2022

시대의 흐름에 따라 정보처리기술이 발전하면서 원자력시설에 대한 사이버위협 가능성이 갈수록 높아지고 있다. 국외는 2000년대에 들어 원자력시설에 대한 사이버 공격 대비가 필요하다는 인식이 늘어났으며, 실질적으로 사이버공격에 대비하기 위해 원전 사이버보안 규제 체계를 마련하기 시작했다. 국내에서는 사이버위협에 대비하기 위해 2013년과 2014년에 원자력시설 등의 방호 및 방사능 방재 대책법, 시행령 및 시행규칙의 개정 및 방사능방재법 관련 고시를 개정하였다. 그리고 2015년에 국내 원자력사업자는 개정된 법령에 따라 시설별 정보시스템 보안규정을 마련하여 원자력안전위원회로부터 7단계로 나눠진 정보시스템 보안규정 이행계획을 승인받게 되었다. 2019년에는 단계별 이행에 대한 특별검사가 완료되었고, 2019년이 지난 이후부터는 정기검사를 통해 사업자의 사이버보안 체계를 지속적으로 점검해오고 있다. 본 논문에서는 지속적으로 발전하는 원자력시설에 대한 사이버위협에 대응하기 위해 꾸준히 개정되는 원자력 시설 사이버보안 체계 점검에 적합하도록 개선된 훈련을 구축하기 위한 몇 가지 방안에 대해 제시한다.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.40-47
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• 2013

The capacity factor of nuclear power plant in Korea is the highest level in the world. However, the integrity assessment of nuclear power plant is depended on foreign country. Especially, most eddy current testing systems for inspecting steam generators in nuclear power plant are currently imported from USA, Canada, and so on. Therefore, the eddy current testing system can react more active and adaptive from economic and managerial standpoint for actual nuclear power plants in Korea is required. In this paper, an eddy current testing system for inspecting steam generators in nuclear power plants is introduced. Frequency generator, analog circuit, analog digital converter circuit, and digital control circuit are composed in eddy current testing system. A benchmarking of acquisition system and acquisition software, eddynet 11i made by Zetec, and modifications are carried out based on the test environment of Korea nuclear power plants. Finally, all eddy current apparatus are integrated to inspect steam generator tubes in nuclear power plants.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.949-953
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• 2019

The burnup equation of nuclides is one of the most equations in nuclear reactor physics, which is generally coupled with transport calculations. The burnup equation describes the variation of the nuclides with time. Because of its very stiffness and the need for large time step, this equation is solved by special methods, for example transmutation trajectory analysis (TTA) or the matrix exponential methods where the matrix exponential is approximated by CRAM. However, TTA or CRAM functions well when the flux is constant. In this work, a new method is proposed when the flux changes. It's an improved method compared to TTA or CRAM. Furtherly, this new method is based on TTA or CRAM, and it is more accurate than them. The accuracy and efficiency of this method are investigated. Several cases are used and the results show the accuracy and efficiency of this method are great.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.437-441
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• 2022

In this study, we developed a neutron time-of-flight (nTOF) measurement system for a 1.7-MV tandem proton accelerator with a target covered with 300-nm-thick lithium (Li) layer. With implementation of beam chopping module after its ion source, the accelerator is configured to operate in pulsed-beam mode with a pulse width <50 ns at 20-kHz repetition rate. This enables the gamma flash-type nTOF measurement system to identify the neutron generated with 3-MeV proton beam energy. The nTOF system consists of a 30" cylindrical NaI(Tl) and four stilbene scintillation detectors. The NaI(Tl) scintillator is placed 50 cm from the Li target to measure the time of beam irradiation on the target, and the stilbene detectors are placed 2 and 2.4 m away to measure nTOF at each location. The nTOF system successfully measured the generated neutron energy at irradiated proton energies of 2.6 and 3.0 MeV with an average energy resolution of 15%.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4335-4349
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• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.