• 예술인문사회 융합 멀티미디어 논문지
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• 제8권4호
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• pp.285-296
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• 2018

지난 몇년간 발생된 수많은 정보시스템 해킹중 국가적 위험을 불러일으킬만한 사이버공격은 핵시설과 원전에 대한 사이버공격일것이다. 그중 대표적인 것이 이란 핵시설 스턱스넷 공격과 한국의 한수원 사이버위협이라고 할 수 있다. 전자는 직접적인 사이버공격으로 인해 원전이 멈출 수 있다는 것을 보여주었고, 후자는 사이버해킹 위협만으로도 국민을 공포에 떨게 할 수 있다는 것을 보여주었다. 이 사건들 이후 산업 제어시설에 대한 사이버공격 위험성이 알려져서 보안이 강화되기 시작하였다. 원전 발전소도 이전의 소극적인 인터넷과 분리된 네트워크로 안전하게 운영된다는 개념에서 벗어나서 악성코드 등 사이버테러로 인해 공격을 받을 수 있다는 개념으로 변했다. 두 가지 개념의 차이점은 제어시설도 사이버공격으로 침해될 수 있다는 가능성에 기초하여 대응하기 위한 전략으로 세워졌다는 것이다. 그 점에서 미국은 이미 제어시설에 대한 보안프레임워크를 설정하여 대비하고 있다. 본 논문에서는 제어시설에 대한 사이버보안 공격사례 및 공격 시나리오를 식별하고, 각 시나리오별로 위험요소에 대한 대처방안을 분석한다. 이를 통해 국내 보안프레임워크 설계시 참조해야할 사항을 식별하여 보안성을 강화하고 설계시 안정성을 확보하고자 한다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• 한국방사선학회논문지
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• 제6권1호
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• pp.79-82
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• 2012

요오드는 원자력 시설에서 사고가 발생할 경우 방사선 피폭을 검토할 때 고려해야 할 중요한 핵종 중 하나이다. 그러므로 체르노빌 사고 시 대기 중에는 유기물 형태의 요오드가 비유기물 형태의 요오드보다 많이 관찰되었다. 본 연구에서는 시료의 양 및 측정시간에 변화를 주었으며, 또한 $^{131}I$ 액체선원을 사용하여 증류수에 희석한 시료 및 다시마를 함께 섞은 시료를 이용하여 검출하한치를 측정 분석하였다. 방사능농도 하한치에 들어 인체에는 무해함을 확인 할 수 있었다. $^{131}I$선원의 시간이 흐를수록 카운트가 줄어듦을 알 수 있었다. 반감기를 계산해본 결과 7~9사이의 결과를 얻었고, $^{131}I$를 혼합한 시료의 경우 최고 7일이 지난 후에는 초기 조건에서 반으로 감소한다는 것을 알 수 있었다.

• 분석과학
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• 제13권6호
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• pp.693-698
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• 2000

본 일련의 실험은 가동중인 핵시설 주위의 자연환경내 방사성탄소 농도준위 변화의 간접적인 추적을 통하여 체계적이고 장기간에 걸친 환경감시 목적으로 수행되었다. 나무 나이테 분석을 이용한 방사성탄소 농도 측정 결과는 핵시설 가동 후 농도 준위가 증가한 것을 나타내었으며, 그 변화는 발전량과 밀접한 상관관계가 있음을 알 수 있었다. 한편 섬유소 처리를 통한 안정 동위원소비, ${\delta}^{13}C$을 측정한 결과는 -30‰을 나타내었으며, 이 값은 수동법 및 능동법으로 채취한 대기 시료중의 $^{13}C$값 -17‰ 및 -8‰과는 매우 다른 결과를 나타내었다. 이런 차이는 광합성에 의한 동위원소 분별효과라고 가정할 수 있으나, 이 문제는 심도있는 연구가 필요하다.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.941-950
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• 2013

The Korean Nuclear-Hydrogen Technology Development (NHTD) Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about $1030^{\circ}C$ in BOC (Beginning of Cycle) during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about $1.7{\times}10^{21}\;n/cm^2$, PIE (Post-Irradiation Examination) of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility). This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2077-2083
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• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.516-525
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• 1994

TMI 및 Chernobyl 사고이후 향후 원전에 대한 안전성 향상을 강화하기위해 개량형 원전에 대해 여러가지 피동형 안전설비가 제안되고 있다. 피동냉각계통의 타당성을 검증하고 상세 설계자료를 제공하기 위해, 본 연구는 웨스팅하우스사의 AP-600 피동격납용기와 같은 한쪽 가열면을 갖는 폐쇄유로에 대한 공기 유입구 위치 및 외부영향이 자연순환 공기유량에 미치는 영향과 자연 및 강제대류하에서 대류열전달계수를 조사하였다. 본 실험은 AP-600 격납용기를 1/26로 축소한 segment 유형의 실험장치를 토대로 수행되었다. 자연 및 강제대류 조건하의 공기유로내 특정 위치에서 공기의 속도 및 온도를 측정하였다. 실험결과는 일차원 단순 모델과 비교하였으며, 좋은 일치점을 보였다.

• 방사선산업학회지
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• 제18권1호
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• pp.89-93
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• 2024

Accidents at nuclear facilities and nuclear power plants led to leaks of large amounts of radioactive substances. Of the various radioactive nuclides released, ¹³⁷Cs are radioactive substances generated during the fission of uranium. Therefore, due to the high fission yield (6.09%), strong gamma rays, and a relatively long half-life (30 years), a rapid and efficient removal method and a study of adsorbents are needed. Accordingly, an adsorbent was prepared using Prussian blue (PB), a material that selectively adsorbs radioactive cesium. As a result of evaluating the adsorption performance with the prepared adsorbent, it was confirmed that 82% of the removal efficiency was obtained, and most of the cesium was rapidly adsorbed within 10 to 15 minutes. The purpose of this study was to adsorb cesium using the Prussian blue alginate bead and to compare the change in detection efficiency according to the amount of adsorbent added for quantitative evaluation. However, in this case, it is difficult to determine the detection efficiency using a standard source with the same conditions as the measurement sample, so the efficiency change of the HPGe detector according to the different heights of Prussian blue was calculated through MCNP simulation using certified standard materials (1 L, Marinelli beaker) for radioactivity measurement. It is expected to derive a relational equation that can calculate detection efficiency through an efficiency curve according to the volume of Prussian blue, quantitatively evaluate the activity at the same time as the adsorption of radioactive nuclides in actual contaminated water and use it in the field of nuclear facility operation and dismantling in the future.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• 한국산학기술학회논문지
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• 제20권5호
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• pp.13-19
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• 2019

후쿠시마 원전사고 이후 비상 배터리 시설에 대한 안전성 강화가 요구되면서 리튬 폴리머 배터리를 적용한 새로운 전원 공급장치가 국내에서 세계 최초로 제안되었다. 그러나 제안된 기술을 현장에 적용하기 위해서는 배터리 장치가 설치되는 랙 시스템의 내진 안전성이 요구된다. 본 연구에서는 세계 최초로 72시간 용량 확보를 위해 개발 된 리튬폴리머 배터리 장치를 대상으로 지진 발생 시 전원장치의 안전성을 확보하기 위해 설계 된 전원장치 스트링 및 랙 프레임에 대한 내진성능을 평가하고자 하였다. 실험 결과 1) 단위 랙 시스템의 공진대역은 9 Hz로서 지진하중 전 후의 고유진동수가 변하지 않음에 따라 설계지진하중에 대한 부재 및 부재간의 연결부에 대한 안전성을 확인할 수 있었다. 원전 설계기준 OBE와 SSE에서의 가속도 응답 결과 2) 스트링 제작에 의한 진동 저감 효과가 약 20%정도 보였으며, 3) OBE, SSE 조건에서의 내진시험 결과 랙 프레임 시스템은 설계지진에 대해 안전한 것으로 나타났다. 따라서 본 연구에서 제시한 랙 시스템은 요구 지진력에 대한 구조적 건전성이 입증되었으므로 원전 시설에 적용이 가능한 것으로 나타났다.