• 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.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3188-3198
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• 2024

Efficient and secure decommissioning of nuclear facilities demands advanced technologies. In this context, gamma-ray detection and imaging are crucial in identifying radioactive hotspots and monitoring radiation levels. Our study is dedicated to developing a gamma-ray detection system tailored for integration into robotic platforms for nuclear decommissioning, offering a safe and automated solution for this intricate task and ensuring the safety of human operators by mitigating radiation exposure and streamlining hotspot localization. Our approach integrates a Compton camera based 3D reconstruction algorithm with a single Timepix3 detector. This eliminates the need for a second detector and significantly reduces system weight and cost. Additionally, combining a 3D camera with the setup enhances hotspot visualization and interpretation, rendering it an ideal solution for practical nuclear decommissioning applications. In a proof-of-concept measurement utilizing a 137Cs source, our system accurately localized and visualized the source in 3D with an angular error of 1° and estimated the activity with a 3% relative error. This promising result underscores the system's potential for deployment in real-world decommissioning settings. Future endeavors will expand the technology's applications in authentic decommissioning scenarios and optimize its integration with robotic platforms. The outcomes of our study contribute to heightened safety and accuracy for nuclear decommissioning works through the advancement of cost-effective and efficient gamma-ray detection systems.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.444-452
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• 2019

To assess the availability of a nuclear power plant's dynamic systems, it is necessary to consider the impact of dynamic interactions, such as components, software, and operating processes. However, there is currently no simple, easy-to-use tool for assessing the availability of these dynamic systems. The existing method, such as Markov chains, derives an accurate solution but has difficulty in modeling the system. When using conventional fault trees, the reliability of a system with dynamic characteristics cannot be evaluated accurately because the fault trees consider reliability of a specific operating configuration of the system. The dynamic reliability graph with general gates (DRGGG) allows an intuitive modeling similar to the actual system configuration, which can reduce the human errors that can occur during modeling of the target system. However, because the current DRGGG is able to evaluate the dynamic system in terms of only reliability without repair, a new evaluation method that can calculate the availability of the dynamic system with repair is proposed through this study. The proposed method extends the DRGGG by adding the repair condition to the dynamic gates. As a result of comparing the proposed method with Markov chains regarding a simple verification model, it is confirmed that the quantified value converges to the solution.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1887-1894
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• 2024

Fatigue strength of the structural materials of lead-cooled fast reactors (LFRs) and accelerator-driven systems (ADS) may be degraded in liquid metal (Lead or lead-bismuth eutectic (LBE)) environments. The fatigue crack growth (FCG) data of structural materials in liquid LBE are necessary for damage tolerance design, safety assessment and life management of key equipment. A novel monitoring system for fatigue crack length was designed on the compliance method and the monitor technology of crack opening displacement (COD) of CT specimens by the linear variable differential transformers (LVDT) system. It can be used to predict the crack length by monitoring the COD of CT specimens in harsh high-temperature liquid LBE using a LVDT system. The prediction accuracy of this system was verified by FCG experiments in room temperature air and liquid LBE at 150, 250 and 350 ℃. The first results obtained in the FCG test for T91 steel in liquid LBE at 350 ℃ are presented.

• 대한핵의학회지
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• 제19권2호
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• pp.101-103
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• 1985

Azygos system, one of the main collateral vessels which communicates superior vena cava with inferior vena cava, is well visualized by Xray angiography. This system is rarely demonstrated by radioisotope study. We report a case whose azygos system was shown during $^{99m}Tc-DTPA$ renal scanning.