• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.167-167
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.179-179
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• 2002

• Journal of Energy Engineering
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• v.23 no.4
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• pp.19-30
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• 2014

The leading countries in nuclear technology development are concentrating their efforts on the development of Sodium-cooled Fast Reactor, which is one of the Generation-IV nuclear reactor systems characterized by a sustainability, an enhanced safety, proliferation resistance, and improved economics. Especially, the Republic of Korea is developing a Sodium-cooled Fast Reactor equipped with metallic-fuel. This type of fast reactor has superior inherent safety and passive safety characteristics. Further, sodium-cooled fast reactors enable the reuse of spent fuel and the closing of fuel cycle, thus, it increases the sustainability of nuclear energy. Many countries are planning the deployment of sodium-cooled fast reactors before 2050 in their energy mix.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Journal of the Korean Society of Marine Environment & Safety
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• v.23 no.3
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• pp.224-230
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• 2017

The purpose of this study is to classify the deficient abilities of seafarers into SRK (Skill, Rule, and Knowledge) and analyze and identify the SRK by the type of accident and ship. Experimental data used the SRK cumulative frequency for 1,606 marine accident records and two-way ANOVA and t-test were used for the analysis tools. The results of two-way ANOVA showed that it is possible to identify the deficient abilities by using the cumulative frequency of SRK in both accident and ship types. As a result of the t-test, the adoption of the null hypothesis (H=0) that the mean of two pairs is equal and the rejection of the null hypothesis (H=1) were 29.2 % and 70.8 %, respectively. For the ship type, H=0 is 33.3 % and H=1 is 66.7 %. Through this study, it was found that about 70 % of the deficient abilities of seafarers inherent in maritime accidents can be identified using the proposed method.

• Journal of the Korean Society of Marine Environment & Safety
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• v.26 no.3
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• pp.254-261
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• 2020

Welding is the most convenient method for fabricating steel materials to build ships and of shore structures. However, welding using high heat processes inevitably produces welding displacements on welded structures. To mitigate these, heavy industries introduce various welding techniques such as back-step welding and skip-step welding. These techniques effect on the change of the distribution of high heat on welded structures, leading to a reduction of welding displacements. In the present study, various cases using different and newly introduced welding techniques are numerically simulated to ascertain the most efficient technique to minimize welding displacements. A numerical simulation using a finite element method based on the inherent strain, interface element and multi-point constraint function is introduced herein. Based on several simulation results, the optimal welding technique for minimizing welding displacements to build a general ship grillage structure is finally proposed.

• Journal of Korea Society of Digital Industry and Information Management
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• v.9 no.4
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• pp.81-94
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• 2013

Recently, the dramatical increasement of personal information infringement makes the government strongly enforce the laws. The Key-point of law enforcement is the DB encryption. Nevertheless, DB encryption is the one of the hardest thing in the organization's security measures. The purpose of this paper is suggesting alternative means of residence numbers and showing the possibility of indicator usage for safety measures. This research suggested the best ways to make a decision through a before and after comparison of the DB encryption cost of the inherent identification number elimination in "Social Welfare consolidation service system". When this research result was applied in "Happiness-e-Um system", we found that the alternative means are more efficient than the residence number for encryption cost, system revision time and so on.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1372-1386
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• 2018

Despite recent advances in multi-hazard analysis, the complexity and inherent nature of such problems make quantification of the landslide effect in a probabilistic safety assessment (PSA) of NPPs challenging. Therefore, in this paper, a practical approach was presented for performing an earthquake-induced landslide PSA for NPPs subject to seismic hazard. To demonstrate the effectiveness of the proposed approach, it was applied to Korean typical NPP in Korea as a numerical example. The assessment result revealed the quantitative probabilistic effects of peripheral slope failure and subsequent run-out effect on the risk of core damage frequency (CDF) of a NPP during the earthquake event. Parametric studies were conducted to demonstrate how parameters for slope, and physical relation between the slope and NPP, changed the CDF risk of the NPP. Finally, based on these results, the effective strategies were suggested to mitigate the CDF risk to the NPP resulting from the vulnerabilities inherent in adjacent slopes. The proposed approach can be expected to provide an effective framework for performing the earthquake-induced landslide PSA and decision support to increase NPP safety.