• Journal of Radiation Protection and Research
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• 제25권3호
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• pp.133-135
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• 2000

• 대한핵의학회:학술대회논문집
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• 대한핵의학회 2003년도 제42차 춘계학술대회 및 총회
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• pp.19-21
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• 2003

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.305-306
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• 2017

• 설비공학논문집
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• 제29권12호
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• 한국재난정보학회 논문집
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• 제17권3호
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• pp.500-511
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• 2021

연구목적: 후쿠시마 원자력 발전소 폭발사고 발생 당시의 재난상황, 주민보호를 위한 비상발령 상황, 주민 피난 상황에 대한 조사와 더불어 당시 주민보호 시스템에 있어서 제기된 문제점 및 조치사항에 대한 조사를 수행함으로서, 국내 원자력 재해를 대비한 주민 피난 대책 수립 과정에서 검토해야만 하는 시사점을 도출하고자 한다. 연구방법: 동일본 대지진 직후부터 현재까지 일본의 국가, 국회, 지자체 및 관련기관으로부터 발행된 보고서를 중심으로 문헌조사를 수행하였다. 연구결과: 후쿠시마 원자력발전소 폭발사고 발생 당시의 주민의 피난과정에서 도출된 문제점과 대응방안에 대한 조사 결과를 통하여 국내의 방사능 재해 대책 수립과정에 있어서 검토해야만 하는 사항을 도출하였다. 결론: 검토 사항을 크게 4가지로 분류하였으며 각 분류에 따른 상세 검토사항을 제시하였다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.327-334
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• 2017

Common-cause failure (CCF) events can significantly impact the availability of safety systems of nuclear power plants. For this reason, the International Common Cause Data Exchange (ICDE) project was initiated by several countries in 1994. Since 1997 it has been operated within the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) framework and has successfully been operated over six consecutive terms (the current term being 2015-2017). The ICDE project allows multiple countries to collaborate and exchange CCF data to enhance the quality of risk analyses, which include CCF modeling. As CCF events are typically rare, most countries do not experience enough CCF events to perform meaningful analyses. Data combined from several countries, however, have yielded sufficient data for more rigorous analyses. The ICDE project has meanwhile published 11 reports on the collection and analysis of CCF events of specific component types (centrifugal pumps, emergency diesel generators, motor operated valves, safety and relief valves, check valves, circuit breakers, level measurement, control rod drive assemblies, and heat exchangers) and two topical reports. This paper presents recent activities and lessons learnt from the data collection and the results of topical analysis on emergency diesel generator CCF impacting entire exposed population.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1307-1313
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• 2019

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.

• Journal of Radiation Protection and Research
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• 제27권3호
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• pp.189-198
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• 2002

방사선 비상사고시 예상되는 주민행동특성 조사 및 교통분석을 통해 실제적인 가정에 기초한 울진원전 비상계획구역내 주민들에 대한 소개시간평가를 수행하였다. 본 연구에서 소개시간은 주민통보, 소개준비 및 차량소개시간으로 구성되었다. 소개대상인구는 비상계획구역내 인구밀도 행정구역 및 일시체류인구 등을 고려해 4개의 그룹으로 분류하였다. 주민행동특성 조사를 위해 비상계획구역내 200가구에 대해 설문조사를 실시하였으며, 설문조사에는 가상사고상황을 설명하는 시나리오를 포함하여, 거주지, 소개준비 소요시간, 소개시 교통수단, 대피장소, 소개방향 등에 대한 질의를 포함하였다. 계산된 소개시작 시간분포 및 미시적 교통분석모델인 CORSIM을 이용하여 도로상에서 소개하는 각 차량들의 거동을 모사하였다. 본 연구결과에서 모든 소개대상차량이 비상계획구역 외부로 소개하는 데 있어서는 밤보다는 낮에 소개하는 경우에 더 오랜 시간이 소요되며 반면에 교차로에서의 지체시간은 낮보다는 밤이 더 장시간 지체되는 것을 확인할 수 있었다. 이것은 차량소개 시작분포에 의한 영향에 기인하는 것으로 분석되었다. CORSIM 모델이 비상사고시 나타날 수 있는 혼잡한 교통현상을 적절히 모사할 수 있는 가를 검증하기 위해 오전 출근시간대에 울진원전 주변의 신호등이 없는 교차로에서 Benchmark Test를 수행하였다. 이 시험에서 CORSIM 모델의 예측치는 관찰된 통과차량 수와 잘 일치하여 본 연구목적을 만족시키고 있음을 확인할 수 있었다.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2379-2386
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• 2020

This paper presents the results of the first intercomparison exercise performed by the Korea retrospective dosimetry (KREDOS) working group using electron paramagnetic resonance (EPR) spectroscopy. The intercomparison employed the alanine dosimeter, which is commonly used as the standard dosimeter in EPR methods. Four laboratories participated in the dose assessment of blind samples, and one laboratory carried out irradiation of blind samples. Two types of alanine dosimeters (Bruker and Magnettech) with different geometries were used. Both dosimeters were blindly irradiated at three dose levels (0.60, 2.70, and 8.00 Gy) and four samples per dose were distributed to the participating laboratories. Assessments of blind doses by the laboratories were performed using their own measurement protocols. One laboratory did not participate in the measurements of Magnettech alanine dosimeter samples. Intercomparison results were analyzed by calculating the relative bias, E_n value, and z-score. The results reported by participating laboratories were overall satisfactory for doses of 2.70 and 8.00 Gy but were considerably overestimated with a relative bias range of 10-95% for 0.60 Gy, which is lower than the minimum detectable dose (MDD) of the alanine dosimeter. After the first intercomparison, participating laboratories are working to improve their alanine-EPR dosimetry systems through continuous meetings and are preparing a second intercomparison exercise for other materials.