• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.525-536
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• 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.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.64 no.2
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• pp.254-259
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• 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.

• Proceedings of the KSR Conference
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• 2007.05a
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• pp.968-973
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• 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
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• v.47 no.2
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• pp.187-195
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• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.347-357
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• 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.

• Journal of Environmental Health Sciences
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• v.50 no.1
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• pp.1-5
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• 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
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• v.38 no.2
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• pp.60-67
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• 2013

A diffusion coefficient is an important parameter in the prediction of atmospheric dispersion using a Gaussian plume model, and its modelling approach varies. In this study, dispersion coefficients recommended by the U. S. Nuclear Regulatory Commission's (U. S. NRC's) regulatory guide and the Canadian Nuclear Safety Commission's (CNSC's) regulatory guide, and used in probabilistic accident consequence analysis codes MACCS and MACCS2 have been investigated. Based on the atmospheric dispersion model for a hypothetical accidental release recommended by the U. S. NRC, its influence to atmospheric dispersion factor was discussed. It was found that diffusion coefficients are basically predicted from a Pasquill- Gifford curve, but various curve fitting equations are recommended or used. A lateral dispersion coefficient is corrected with consideration for the additional spread due to plume meandering in all models, however its modelling approach showed a distinctive difference. Moreover, a vertical dispersion coefficient is corrected with consideration for the additional plume spread due to surface roughness in all models, except for the U. S. NRC's recommendation. For a specified surface roughness, the atmospheric dispersion factors showed differences up to approximately 4 times depending on the modelling approach of a dispersion coefficient. For the same model, the atmospheric dispersion factors showed differences by 2 to 3 times depending on surface roughness.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.235-243
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• 2016

In the near future, many countries, including the Republic of Korea, will face a significant increase in low level radioactive waste (LLW) from nuclear power plant decommissioning. The purpose of this paper is to look at blending as a method for enhancing disposal options for low-level radioactive waste from the decommissioning of nuclear reactors. The 2007 U.S. Nuclear Regulatory Commission strategic assessment of the status of the U.S. LLW program identified the need to move to a risk-informed and performance-based regulatory approach for managing LLW. The strategic assessment identified blending waste of varying radionuclide concentrations as a potential means of enhancing options for LLW disposal. The NRC's position is that concentration averaging or blending can be performed in a way that does not diminish the overall safety of LLW disposal. The revised regulatory requirements for blending LLW are presented in the revised NRC Branch Technical Position for Concentration Averaging and Encapsulation (CA BTP 2015). The changes to the CA BTP that are the most significant for NPP operation, maintenance and decommissioning are reviewed in this paper and a potential application is identified for decommissioning waste in Korea. By far the largest volume of LLW from NPPs will come from decommissioning rather than operation. The large volumes in decommissioning present an opportunity for significant gains in disposal efficiency from blending and concentration averaging. The application of concentration averaging waste from a reactor bio-shield is also presented.

• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.127-134
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• 1980

Shielding problems associated with neutron streaming through the reactor vessel cavity of pressurized water reactors are discussed to a certain extent with the actual examples in the currently operating reactors. Various remedial techniques are proposed herein to mitigate the tedious neutron streaming phenomena including piling up in heaps of temporary boron-containing bags and the installation of permanent shield structure making use of a certain refractory materials. In conclusion, optimum cavity shielding design concepts are presented with special emphasis on such major factors as the identification of major neutron streaming path, selection of necessary shielding materials with acceptable constraints, detailed design characteristics and physical configuration as well as the formulation of dependable mathematical tools to predict the final outcome of each design concept proposed in the context.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.51-59
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• 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.