• Nuclear Engineering and Technology
• /
• 제37권6호
• /
• pp.525-536
• /
• 2005

The interfacial area transport equation dynamically models the changes in interfacial structures along the flow field by mechanistically modeling the creation and destruction of dispersed phase. Hence, when employed in the numerical thermal-hydraulic system analysis codes, it eliminates artificial bifurcations stemming from the use of the static flow regime transition criteria. Accounting for the substantial differences in the transport mechanism for various sizes of bubbles, the transport equation is formulated for two characteristic groups of bubbles. The group 1 equation describes the transport of small-dispersed bubbles, whereas the group 2 equation describes the transport of large cap, slug or chum-turbulent bubbles. To evaluate the feasibility and reliability of interfacial area transport equation available at present, it is benchmarked by an extensive database established in various two-phase flow configurations spanning from bubbly to chum-turbulent flow regimes. The geometrical effect in interfacial area transport is examined by the data acquired in vertical fir-water two-phase flow through round pipes of various sizes and a confined flow duct, and by those acquired In vertical co-current downward air-water two-phase flow through round pipes of two different sizes.

• 전기학회논문지
• /
• 제64권2호
• /
• pp.254-259
• /
• 2015

On January 30, 2012 an auxiliary component of Byron Unit 2 was tripped on bus under voltage. The cause of the event was the failure of the C-phase insulator track for the Unit 2 station auxiliary transformer(SAT) revenue metering transformer. In addition to this event, other events have occurred at other plants resulting in an open-phase condition.[1] Therefore, Nuclear Regulatory Commission(NRC) has requested that not only nuclear power plant(NPP) operating company but also its Design Certification(DC) applicant have to prepare open-phase detecting system in their operating plants and design document. In this paper, various open-phase conditions are simulated in NPP using Electromagnetic Transient Program(EMTP) and Atpdraw, and open-phase condition detecting system is proposed for Main Transformer(MT), Unit Auxiliary Transformer(UAT) and SAT connected power line in NPP.

• 한국철도학회:학술대회논문집
• /
• 한국철도학회 2007년도 춘계학술대회 논문집
• /
• pp.968-973
• /
• 2007

The systems such as the railway control system, satellite control system and nuclear power plant control system are the safety critical systems because the failure of them could lead to risk significant events. These softwares of digital systems must follow the life cycle process from the beginning of software development to guarantee their safety and reliability. The NRC(Nuclear Regulatory Commission) Reg Guide of nuclear fields, the RTCA/DO-178B standard which is used to acquire the certification for software in industrial aero field in European Union and United State, the DEF STAN 00-55 standard for the safety of electronic weapon in England, the IEC 601-1-4 for medical equipment and the IEC 62279 for railway system recommended the development life cycle. This paper introduces the development process and compares each other. Also it indicates applicable development criteria for the software of systems related to railway fields and describes the detailed procedure of development criteria. We describe the procedure to make the software development criteria in nuclear filed. For the software development related to railways, the process from plan phase to maintenance phase must be satisfied. The safety and reliability is guaranteed through these standards.

• Nuclear Engineering and Technology
• /
• 제47권2호
• /
• pp.187-195
• /
• 2015

Human-induced initiating events, also called Category B actions in human reliability analysis, are operator actions that may lead directly to initiating events. Most conventional probabilistic safety analyses typically assume that the frequency of initiating events also includes the probability of human-induced initiating events. However, some regulatory documents require Category B actions to be specifically analyzed and quantified in probabilistic safety analysis. An explicit modeling of Category B actions could also potentially lead to important insights into human performance in terms of safety. However, there is no standard procedure to identify Category B actions. This paper describes a systematic procedure to identify Category B actions for low power and shutdown conditions. The procedure includes several steps to determine operator actions that may lead to initiating events in the low power and shutdown stages. These steps are the selection of initiating events, the selection of systems or components, the screening of unlikely operating actions, and the quantification of initiating events. The procedure also provides the detailed instruction for each step, such as operator's action, information required, screening rules, and the outputs. Finally, the applicability of the suggested approach is also investigated by application to a plant example.

• 방사성폐기물학회지
• /
• 제21권3호
• /
• pp.347-357
• /
• 2023

United States Nuclear Regulatory Commission (U.S. NRC) specifies regulations on obtaining licenses and describes the technical position on the average waste concentration, also known as Concentration Averaging and Encapsulation Branch Technical Position (CA BTP); CA BTP helps classify blendable waste and discrete items and address concentration averaging. The technical position details are reviewed and compared in a real environment in Korea. A few cases of concentration averaging based on the application of CA BTP to domestic radioactive waste are presented, and the feasibility of the application is assessed. The radioactive waste considered herein does not satisfy the Disposal Concentration Limit (DCL) of the second-phase disposal facility while applying the preliminary classification. However, if CA BTP is applied when the radioactive waste is mixed with other radioactive waste items in a large and heavy container, it can be disposed of at the second-phase disposal facility in Gyeongju Repository. To apply the CA BTP of the U.S. NRC, it is necessary to investigate the safety assessment conditions of the US and Korea.

• 한국환경보건학회지
• /
• 제50권1호
• /
• pp.1-5
• /
• 2024

International Commission on Radiological Protection (ICRP) is an independent international organization that advances the science of radiological protection for the public benefit, particularly by providing recommendations and guidance on all aspects of protection against ionizing radiation. The ICRP is a community of more than 380 globally-recognized experts in radiological protection science, policy, and practice from more than 50 countries. As of January 2024, the ICRP is comprised of a Main Commission, the Scientific Secretariat, four Standing Committees, and 30 Task Groups under the four committees. The ICRP has released well over one hundred publications on all aspects of radiological protection. Most address a particular area within radiological protection, but a handful of the publications, the so-called fundamental recommendations, describe the overall system of radiological protection. The system for radiological protection is based on the current understanding of the science of radiation exposure and its effects along with value judgements. The ICRP offers recommendations to regulatory and advisory agencies and provides advice to management and professional staff with responsibilities for radiological protection. Legislation in most countries adheres closely to ICRP recommendations. The International Atomic Energy Agency's (IAEA) International Basic Safety Standards are based heavily on ICRP recommendations. ICRP recommendations form the core of radiological protection standards, legislation, programs, and practice worldwide.

• Journal of Radiation Protection and Research
• /
• 제38권2호
• /
• pp.60-67
• /
• 2013

가우시안 플륨모델(Gaussian plume model)을 사용한 대기확산의 예측에서 확산계수는 결과에 중요한 영향을 미치는 변수이다. 확산계수의 평가방법은 다양하며, 본 연구에서는 미국 원자력규제위원회(U. S. NRC) 권고 규제지침, 캐나다 원자력안전위원회(CNSC) 권고 규제지침, 확률론적 사고결말해석코드 MACCS와 MACCS2에서 권고 또는 적용하는 방법을 고찰하였다. U. S. NRC에서 권고하는 부지적합성 평가를 위한 가상사고시 대기확산모델을 기반으로 확산계수의 평가방법이 대기확산인자에 미치는 영향을 분석하였다. 확산계수는 Pasquill-Gifford 곡선을 기반으로 각기 다른 연구자들에 의해 얻어진 곡선의 피팅식(curve fitting equations)을 적용 또는 권고하고 있음을 확인하였다. 수평확산계수는 모든 규제지침과 코드에서 플륨의 사행효과를 반영하여 보정하고 있으나 그 적용 방법에 있어서는 차이를 나타냈다. 수직확산계수는 U. S. NRC 권고 규제지침을 제외하고 표면거칠기를 반영하여 보정하고 있다. 특정 표면거칠기에 대해 확산계수의 적용방법에 따라 대기확산인자는 최대 약 4배의 차이를 나타냈다. 표면거칠기는 대기확산인자에 중요한 영향을 나타냈으며, 동일 적용방법에 대해 표면거칠기에 따라 대기확산인자는 약 2~3배의 차이를 나타냈다.

• 방사성폐기물학회지
• /
• 제14권3호
• /
• pp.235-243
• /
• 2016

우리나라를 포함한 많은 국가들에서 향후 원전 해체로 저준위폐기물이 대량으로 발생할 전망이다. 본 논문에서는 미국의 저준위방사성폐기물 처분 관련 규제 기준을 분석하고, 특히 원자력발전소의 운영 및 해체를 포함하는 전주기에서 발생하는 폐기물의 처분 옵션을 확장하는 방안으로 사용되고 있는 저준위방사성폐기물의 블랜딩에 대해 검토하였다. 2007년 미국 NRC는 미국 저준위폐기물 관리 프로그램에 대한 전략분석 결과, 방사선위험도와 성능평가에 기반한 새로운 저준위폐기물 관리 규제의 필요성을 제기하였는데, 특히 방사성핵종 농도가 다른 폐기물의 블랜딩을 처분에 대한 옵션을 다양화할 수 있는 안전한 방안으로 제시하였다. NRC는 블랜딩을 처분에 적합하도록 방사성핵종의 농도가 다른 저준위폐기물을 비교적 균일하게 혼합(mixing)하는 것으로 정의하였다. 2015년 2월 농도 평균과 포장에 대한 NRC BTP의 개정판으로 공표된 블랜딩에 대한 구체적인 기술요건을 분석하였고 국내 해체폐기물에 대한 적용 방안도 예시하였다. 대량으로 발생할 해체폐기물에 대해 블랜딩과 농도평균을 적용하면 처분 효율성을 향상시킬 수 있다. 바이오쉴드 콘크리트에 대한 농도평균 적용에 대해 예시하였다.

• Nuclear Engineering and Technology
• /
• 제12권2호
• /
• pp.127-134
• /
• 1980

최근 가압경수로에서 압력용기 상단주변의 캐비티에 중성자가 새어나오는 사실이 판명되자 이에 대한 차폐문제가 심각하게 논의되기 시작했다. 본 논문에서는 현재 운전중인 원자로에서 이것을 어떻게 해결하고 있는가를 예시하였다. 예를들면 붕소가 들어있는 주머니를 쌓아 올리는 것에서부터 화폐구조물을 영구히 설치하는 것에 이르기까지 여러가지 방법으로 중성자흐름을 막는 대책을 논의하였다. 결론적으로 이 문제해결을 위한 가장 현실적이고 가능한 차폐설계안 몇가지를 제시하였다. 특히 그중에서도 중성자가 어떤 경로로 흘러 나오는가의 규명. 차폐자료의 선정, 차폐설계의 특성과 외형문제 그리고 각 설계안을 검토키 위한 수학적 모델 제시에 역점을 두었다.

• 한국압력기기공학회 논문집
• /
• 제13권2호
• /
• pp.51-59
• /
• 2017

Steam jets ejected from a rupture zone of high energy pipes may cause damage to adjacent structures. This event could lead to more serious accidents in nuclear power plants. Therefore, to prevent serious accidents, high energy pipes of nuclear power plants are designed according to the ANSI / ANS 58.2 technical standard. However, the US Nuclear Regulatory Commission (USNRC) has recently pointed out non-conservatism in existing high energy pipe fracture evaluation methods, and required the assessment of the unsteady load of the jet caused by a potential feedback mechanism as well as the impact range of steam jet, the jet impact loads and the blast wave effects at the initial breakage stage. The potential feedback mechanism refers to a phenomenon in which a vortex formed by impingement jets amplifies vortex itself and induces jet vibration in a shear layer. In this study, CFD methodology using the LES turbulence model is established and numerical analysis is carried out to evaluate the dynamic behavior of impingement jets and the potential feedback mechanism during jet impingement. Obtained results have been compared with an empirical correlation and experiment.