• 광물과 암석
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• 제35권1호
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• pp.25-39
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• 2022

본 연구는 원자력 시설 해체 시 발생되는 저준위 및 극저준위 폐토양, 점토와 산업부산물인 고로슬래그를 이용하여 방사성 폐기물을 안전하게 담지할 수 있는 비소성 시멘트의 제조 가능성을 평가하고 광물·형태학적 분석을 통하여 생성된 반응 물질에 대하여 고찰하였다. 본 연구에서는 (1) 폐토양, 점토 및 고로슬래그의 특성 분석, (2) 폐토양, 점토 및 고로슬래그를 고화재 및 성분조정제로 이용한 원전 해체 폐기물 담지를 위한 비소성 시멘트 제조 및 최적의 배합 비율 도출, (3) 제조된 비소성 시멘트 고화체의 수화반응 생성물질에 대하여 광물·형태학적 분석 등을 수행하였다. 비소성 시멘트 고화체의 광물·형태학적 분석 결과, 폐토양과 점토는 수화반응 생성물이 관측되지 않았으며, 고로슬래그의 경우 고화체의 강도를 발현시킬 수 있는 수화반응생성물질인 calcium silicate hydrate (CSH), 에트링가이트(ettringite)가 생성되는 것을 확인하였다. 폐토양, 점토를 고화재로 이용한 비소성 시멘트의 재령 28일 후 고화체는 최적의 배합 비율에서 약 3 MPa의 강도를 나타내 처분장 인수기준 압축강도인 3.44 MPa를 만족하지 못하는 것을 확인하였다. 그러나, 고로슬래그를 고화재로 이용한 비소성 시멘트는 모든 실험 조건에서 처분장 인수기준 압축강도를 만족하며, 최적의 배합 비율에서는 약 27 MPa로 높게 나타나는 것을 확인할 수 있었다. 이러한 결과를 통하여 비소성 시멘트 고화재로 고로슬래그, 방사성 핵종에 대한 흡착제 역할로 폐토양 및 점토를 이용한다면 방사성 폐기물 처분을 위한 최적의 비소성 시멘트를 제조할 수 있을 것으로 판단된다.

• 한국산학기술학회논문지
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• 제10권12호
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• pp.3748-3754
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• 2009

원자력발전소에서 냉각재상실사고이후 원자로건물집수조 여과기에서 화학적 영향에 의한 수두손실 변화를 관찰하기 위하여 시험장치에서 단기살수조건, 장기살수조건, 및 화학적 영향을 주는 물질이 없는 조건에 대해 30일 동안 종합적인 수두손실 시험을 수행하였다. 시험결과는 수두손실이 살수조건에 따라 노출된 화학적 영향을 주는 물질의 양에 크게 의존함을 보였다. 시험종료후 수거된 침전물의 XRD 분석은 침전물이 주로 인산화합물임을 보였다. 수두손실과 용해된 화학종의 비교결과는 화학적 영향을 주는 물질 중에서 Al과 Zn의 부식이 시험 초기에 높은 수두 손실 증가율의 원인이 됨을 보였다. 금속 시편에 부동피막이 형성된 이후에 수두손실 증가율은 감소하지만 지속적으로 수두손실이 증가하는 현상은 NUKON 및 콘크리트에서 침출반응에 의해 발생하는 Si, Mg, 및 Ca이 침전물을 형성하는 반응에 기인함을 보였다.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.225-236
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• 2012

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.

• 한국지진공학회논문집
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• 제8권5호통권39호
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• pp.79-89
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• 2004

본 연구에서는 원전기기의 내진안전성을 증가시키기 위해 면진장치를 적용한 기기의 진동대 실험을 수행하였다. 원전구조물과 유사한 진동수 특성을 가지는 실험모형을 제작하여 실험에 사용하였으며 구조물 내부의 기기를 모형화 하기 위하여 400kg의 강체를 사용하였다. 탁월주파수 특성이 상이한 3종류 지진파를 이용하여 진동대 실험을 수행하였다. 면진장치로는 마찰진자형 베어링(FPS)을 사용하였다. 입력지진의 최대가속도를 0.1g, 0.2g, 0.25g의 3단계로 변화시키면서 실험을 수행하였고 또한 1방향, 2방향 및 3방향 가진에 의한 거동을 분석하였다. 실험결과 지진파의 연직성분이 FPS의 면진성능에 영향을 미치는 것을 알 수 있었으며 펄스타입의 속도성분이 큰 근거리 지진인 경우 면진효과가 감소하는 것을 알 수 있었다.

• 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.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.246-258
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• 2016

The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH) and finite element method (FEM) coupling method is used to numerically simulate the fluid and structure interaction (FSI) between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs) in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

• 한국안전학회지
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• 제27권6호
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.439-446
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• 2014

An amphibious inspection robot system (hereafter AIROS) is being developed to visually inspect the in-containment refueling storage water tank (hereafter IRWST) strainer in APR1400 instead of a human diver. Four IRWST strainers are located in the IRWST, which is filled with boric acid water. Each strainer has 108 sub-assembly strainer fin modules that should be inspected with the VT-3 method according to Reg. guide 1.82 and the operation manual. AIROS has 6 thrusters for submarine voyage and 4 legs for walking on the top of the strainer. An inverse kinematic algorithm was implemented in the robot controller for exact walking on the top of the IRWST strainer. The IRWST strainer has several top cross braces that are extruded on the top of the strainer, which can be obstacles of walking on the strainer, to maintain the frame of the strainer. Therefore, a robot leg should arrive at the position beside the top cross brace. For this reason, we used an image processing technique to find the top cross brace in the sole camera image. The sole camera image is processed to find the existence of the top cross brace using the cross edge detection algorithm in real time. A 5-DOF robot arm that has multiple camera modules for simultaneous inspection of both sides can penetrate narrow gaps. For intuitive presentation of inspection results and for management of inspection data, inspection images are stored in the control PC with camera angles and positions to synthesize and merge the images. The synthesized images are then mapped in a 3D CAD model of the IRWST strainer with the location information. An IRWST strainer mock-up was fabricated to teach the robot arm scanning and gaiting. It is important to arrive at the designated position for inserting the robot arm into all of the gaps. Exact position control without anchor under the water is not easy. Therefore, we designed the multi leg robot for the role of anchoring and positioning. Quadruped robot design of installing sole cameras was a new approach for the exact and stable position control on the IRWST strainer, unlike a traditional robot for underwater facility inspection. The developed robot will be practically used to enhance the efficiency and reliability of the inspection of nuclear power plant components.

• 한국광학회지
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• 제28권4호
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• pp.166-171
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• 2017

수소 가스는 연소과정에서 오염물질의 배출이 없는 친환경 에너지원이다. 그러나 연소 및 폭발성이 매우 강해 매우 위험한 특징을 갖고 있다. 원자력 발전소의 중대 사고 발생시 핵연료의 산화 과정에서 다량의 수소 가스가 발생하며 원전 격납 건물의 2차 사고의 원인으로 작용함으로 원전의 안전을 확보하기 위하여 수소 가스의 검출 기술은 매우 중요하다. 본 논문은 수소 가스의 원격 계측을 위한 라만 라이다 시스템의 개발에 관한 것이다. 소형의 이동 가능한 라만 라이다 시스템을 설계 및 개발하였으며, 수소 가스의 농도를 정량적으로 계측하기 위한 계측 알고리즘을 개발하였다. 개발된 수소 가스 계측을 위한 라만 라이다 시스템의 수소 가스 검출 능력을 검증하기 위하여 수소 가스의 농도를 조절할 수 있는 가스 챔버를 이용하여 낮에 야외 환경에서 수소 가스 검출 실험을 실시하였다. 그 결과 20미터 거리에서 최소 0.67 Vol.%의 수소 가스 농도의 검출이 가능하였다.

• 한국안전학회지
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• 제29권6호
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• pp.172-178
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• 2014

In terms of an accident management, the cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results since the Three Mile Island (TMI) accident. The objectives of this paper are to explain how to identify the plant response and cope with its vulnerabilities using the probabilistic safety assessment (PSA) quantified results and severe accident database SARDB(Severe Accident Risk Data Bank) based on sequences analysis results. Although PSA has been performed for the Korean Standard Power Plants (KSNPs), and that it considered the necessary sequences for an assessment of the containment integrity. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by a knowledge-based technique, and the expected plant behaviour. The plant model used in this paper is oriented to the cases of loss of coolant accident (LOCA) is be used as a training simulator for a severe accident management.