• Nuclear Engineering and Technology
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• v.38 no.3
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• pp.251-258
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• 2006

This study presents an idea of using analogy concept to the heat transfer studies regarding the HTGR development. Theoretical backgrounds regarding the idea were reviewed. In order to investigate the predictability of a mass transfer system for heat transfer system phenomenology, an electroplating system coupled with a limiting current technique was adopted. Test facilities for laminar forced convection and natural convections under laminar and turbulent conditions were constructed, for which heat transfer correlations are known. The test results showed a close agreement between mass transfer and heat transfer systems, which is an encouraging indication of the validity of the analogy theory and the experimental methodology adopted. This paper shows the potential of the experimental method that validates the little-understood heat transfer phenomena, even in complex geometries such as HTGR.

• Nuclear Engineering and Technology
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• v.9 no.3
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• pp.151-163
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• 1977

Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design Parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1) , currently under construction in Korea aiming at its completion in 1982. An attempt was also made for tile sensitivity analysis of each fuel component; j. e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1015-1021
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• 2016

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.391-402
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• 2009

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2019.05a
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• pp.248-249
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• 2019

In the nuclear industry, in order to keep pace with the 4th industrial revolution era, they are trying to improve the construction and maintenance ability by utilizing the technologies such as digital twin and VR/AR from the construction stage. However, the nuclear 3D CAD model, which is used as the base in the latest technology, is heavy due to a large number of facilities per unit space compared to other industrial companies, and it is difficult to directly incorporate the latest technology into the results of CAD programs for design purposes. In this study, in order to improve usability, we tried to lighten the 3D model. First, we analyze the existing nuclear power plant 3D model and draw out the problems and features. Secondly, we derived the factors to consider when we make the 3D CAD models lightweight.

• Proceedings of the Korea Information Processing Society Conference
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• 2015.04a
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• pp.449-451
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• 2015

When cyber attacks are discovered in nuclear facilities, licensees are required to notify regulatory organizations for quick action. This also helps regulatory organizations to strengthen regulatory capabilities for cyber security. Currently the U.S. issued the final draft rule for Cyber Security Event Notifications. Domestic regulatory activities being at an early stage for cyber security need to implement law for Cyber Security Event Notifications. Since the current laws are focused on the aspect of safety, they are in need of more specific laws for cyber security.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1664-1675
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• 2021

In order to verify the statistical performance of the nonparametric tests used in the MARSSIM approach, all plausible contamination distribution types that can be encountered in a survey area should be investigated. As the first of such investigations, this study aims to perform the verification for normal distribution of the contamination in a survey area by simulating the collection of random samples from it through the Monte Carlo simulation. The results of the simulations conducted for a total of 81 simulation cases showed that Sign test and WRS test both exhibited an excellent statistical performance: 100% for the former and 98.8% for the latter. Therefore, in final status surveys of the MARSSIM approach, a high statistical performance can be expected in applying the nonparametric hypothesis tests to survey areas whose net contamination can be assumed to be normally distributed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.321-338
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• 2022

The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three-dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of ¹³⁷Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

• Proceedings of the Korea Information Processing Society Conference
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• 2020.05a
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• pp.120-122
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• 2020

원전에 대한 사이버위협이 지속됨에 따라 IAEA 및 각국에서는 원전 사이버보안 강화를 위해 노력하고 있다. 그 일환으로 국내에서는 규제기준 KINAC/RS-015를 통해 원전 내 안전·보안·비상대응 기능과 관련된 필수디지털자산에 대한 사이버보안 규제를 수행하고 있으나 원전 사고와 직접적으로 관련된 자산에 대해서는 보다 강화된 보안조치를 적용하여 보안성을 높이고자 한다. 이러한 강화 조치의 하나로 '코드 난독화 적용'이 있으며 이에 대해 상세히 살펴보고자 한다.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2356-2365
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• 2023

Characterization, dismantling and pre-disposal management of irradiated graphite (i-graphite) have an important role in safe decommissioning of several nuclear facilities which used this material as moderator and reflector. In addition to common radiation protection issues, easily volatizing long-lived radionuclides and stored Wigner energy could be released during imprudent retrieval and processing of i-graphite. With this regard, among all cutting technologies, abrasive waterjet (AWJ) can successfully achieve all of the thermo-mechanical and radiation protection objectives. In this work, factorial experiments were designed and systematically conducted to characterize the AWJ processing parameters and the machining capability. Moreover, the limitation of dust production and secondary waste generation has been addressed since they are important aspects for radiation protection and radioactive waste management. The promising results obtained on non-irradiated nuclear graphite blocks demonstrate the applicability of AWJ as a valid technology for optimizing the retrieval, storage, and disposal of such radioactive waste. These activities would benefit from the points of view of safety, management, and costs.