• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• Journal of computational fluids engineering
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• v.10 no.2
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• pp.1-6
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• 2005

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed af inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve meters (12 m) depth of the reactor pool and cooled by the upward flow that the coolant enters the lower inlet of the plenum, rises up through the grid plate and the core channel and comes out from the outlet of chimney. A fission moly guide tube is extended from the reactor core to the top of the reactor chimney for easily loading a fission moly target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by jet flow. This paper describes an analytical analysis that is the study of the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the flow rate, reduced to about fourteen kilogram per second (14 kg/s) from the original flow rate of sixteen point three kilogram per second (16.3 kg/s) did not show the guide tube jet.

• Journal of the Korea Safety Management & Science
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• v.2 no.4
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• pp.91-102
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• 2000

Nowadays, every kind of system is changed so complex and enormous, it is necessary to assure system reliability, product liability and safety. Fault tree analysis(FTA) is a reliability/safety design analysis technique which starts from consideration of system failure effect, referred to as “top event”, and proceeds by determining how these can be caused by single or combined lower level failures or events. So in fault tree analysis, it is important to find the combination of events which affect system failure. Minimal cut sets(MCS) and minimal path sets(MPS) are used in this process. FTA-I computer program is developed which calculates MCS and MPS in terms of Gw-Basic computer language considering Fussell's algorithm. FTA-II computer program which analyzes importance and function cost of VE consists. of five programs as follows : (l) Structural importance of basic event, (2) Structural probability importance of basic event, (3) Structural criticality importance of basic event, (4) Cost-Failure importance of basic event, (5) VE function cost analysis for importance of basic event. In this study, a method of initiation such as failure, function and cost in FTA is suggested, and especially the priority rank which is calculated by computer-aided decision analysis program developed in this study can be used in decision making determining the most important basic event under various conditions. Also the priority rank can be available for the case which selects system component in FMEA analysis.

• International conference on construction engineering and project management
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• 2009.05a
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• pp.83-90
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• 2009

Natural disasters, such as the recent floods in the Midwest, Hurricane Ike in the Gulf coast region (U.S.), and the earthquake in Sichuan (China), cause severe damage to the infrastructure as well as the associated industries and communities that rely on the infrastructure. The estimated damages due to Hurricane Ike in 2008 were a staggering $27 billion, the third worst in U.S. history. In addition, the worst earthquake in three decades in Sichuan resulted in about 90,000 people dead or missing and $20 billion of the estimated loss. A common observation in the analyses of these natural disaster events is the inadequacy of critical infrastructure to withstand the forces of natural calamities and the lack of mitigation strategies when they occur on the part of emergency-related organizations, industries, and communities. If the emergency-related agencies could identify and fortify the vulnerable critical infrastructure in the preparedness stage, the damage and impacts can be significantly reduced. Therefore, it is important to develop a decision support system (DSS) for identifying region-specific mitigation strategies based on the inter-relationships between the infrastructure and associated industries and communities in the affected region. To establish effective mitigation strategies, relevant data were collected from the affected areas with respect to the technical, social, and economic impact levels. The data analysis facilitated identifying the major factors, such as vulnerability, criticality, and severity, for developing a DSS. Customized mitigation strategies that will help agencies prepare, respond, and recover according to the disaster response were suggested.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4759-4775
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• 2022

The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo' based similarity law was employed for the in-vessel, while Ishii's similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood's metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.08a
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• pp.431-431
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• 2012

A compact tokamak reactor concept as a 14 MeV neutron source is desirable from an economic viewpoint for a fusion-driven transmutation reactor. LAR (Low Aspect Ratio) tokamak allows a potential of high "see full txt" operation with high bootstrap current fractions and can be used for a compact fusion neutron source. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components and are constrained to use ITER physics and technology. In a transmutation reactor, the blanket should produce enough tritium for tritium self-sufficiency and the neutron multiplication factor, keff should be less than 0.95 to maintain sub-criticality. The shield should provide sufficient protection for the superconducting toroidal field (TF) coil against radiation damage and heating effects of the fusion neutrons, fission neutrons, and secondary gammas. In this work, characteristics of transmutation reactor based on LAR tokamak is investigated by using the coupled system analysis.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.457-469
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• 1993

JEF-1-based 50-group cross section set for fast reactor calculations was generated using NJOY system. The set was then examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 27 fast critical assemblies. The calculated results using the new set were also compared with those of ENDF/B-IV or-V-based fast set. In general, the JEF-1-based set shows an improvement in predicting measured integral quantities in comparison with the previous set. With a few exceptions, JEF-1 results are comparable to those of ENDF/B-V.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1226-1233
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• 2017

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. The database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

• Journal of Electrical Engineering and Technology
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• v.9 no.4
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• pp.1248-1257
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• 2014

Overhead transmission lines are crucial components in power transmission systems. Well-designed maintenance strategy for overhead lines is required for power utilities to minimize operating costs, while improving the reliability of the power system. This paper presents a maintenance priority index (MPI) of overhead lines for a reliability centered approach. Proposed maintenance strategy is composed of a state index and importance indices, taking into account a transmission condition and importance in system reliability, respectively. The state index is used to determine the condition of overhead lines. On the other hand, the proposed importance indices indicate their criticality analysis in transmission system, by using a load effect index (LEI) and failure effect index (FEI). The proposed maintenance method using the MPI has been tested on an IEEE 9-bus system, and a numerical result demonstrates that our strategy is more cost effective than traditional maintenance strategies.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.63-72
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• 2014

Pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems in Korea. The Korea Atomic Energy Research Institute has been studying pyroprocess technology, and the conceptual design of an engineering-scale pyroprocess facility, called the Advanced Fuel Cycle (AFC) facility, has been performed on the basis of a 10tHM throughput per year. In this paper, the concept of the AFC facility was introduced, and its safety evaluations were performed. For the safety evaluations, anticipated accident events were selected, and environmental safety analyses were conducted for the safety of the public and workers. In addition, basic radiation shielding safety analyses and criticality safety analyses were conducted. These preliminary safety studies will be used to specify the concept of safety systems for pyroprocess facilities, and to establish safety design policies and advance more definite safety designs.