• Journal of the Korean Society of Systems Engineering
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• v.12 no.2
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• pp.101-114
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• 2016

A model-based systems engineering (MBSE) approach to implementing hardware-based network cybersecurity controls for APR1400 non-safety data network is presented in this work. The proposed design was developed by implementing packet filtering and deep packet inspection functions to control the unauthorized traffic and malicious contents. Denial-of-Service (DoS) attack was considered as a potential cybersecurity issue that may threaten the data availability and integrity of DCS gateway servers. Logical design architecture was developed to simulate the behavior of functions flow. HDL-based physical architecture was modelled and simulated using Xilinx ISE software to verify the design functionality. For effective modelling process, enhanced function flow block diagrams (EFFBDs) and schematic design based on FPGA technology were together developed and simulated to verify the performance and functional requirements of network security controls. Both logical and physical design architectures verified that hardware-based cybersecurity controls are capable to maintain the data availability and integrity. Further works focus on implementing the schematic design to an FPGA platform to accomplish the design verification and validation processes.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.05a
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• pp.52-53
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• 2014

The Work Breakdown Structure (WBS) is a primary tool which provides a framework that defines clear scope of all deliverables throughout the project life cycle. Once the WBS is established in projects, it should allow project team members to measure and manage work performances by the WBS; further, it should provide a reference point when any work scope needs to be redefined. Based on the project information in the Progress and Performance Measurement System (PPMS) of UAE's Barakha Nuclear Power Plant (BNPP) projects, an attempt was made to develop a new WBS which provides hierarchical and systematical decomposition of the total work scope of NPP construction projects while avoiding from the preexistence concept in Korean NPP projects that the WBS is a combination of Physical Breakdown Structure (PBS) and Functional Breakdown Structure (FBS). The unique features of the new WBS are as follows: (1) defined the definition of each level of the WBS, (2) subdivided the WBS into 5 hierarchical levels, and (3) adopted globally used general coding structure. The new WBS provides a basic hierarchical structure for the project scope and can be used as a basic tool for schedule control, performance measurement, project status monitoring, and communication among project participants. In addition, by putting the Work Package (WP) under the WBS, the Earned Value Management System (EVMS) per WP can be utilized for the project.

• KDI Journal of Economic Policy
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• v.40 no.1
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• pp.67-83
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• 2018

We have hypothesized that nuclear risk is significantly inversely related to the distance from residences to nuclear power plants and that the level of life satisfaction of residents therefore increases with the distance. We empirically explore the relationship between Ulsan citizens' life satisfaction levels and the distance between their residences and the Kori and Wolsong nuclear power plants (NPP) based on the life satisfaction approach (LSA). The dataset we used covers only Ulsan citizens from the biennial Ulsan Statistics on Citizen's Living Condition and Consciousness of 2014 and 2016. Controlling for micro-variables such as education, work satisfaction, gender, marital status, and expenditures, we found a statistically significant relationship between life satisfaction and the distance between the residences and the nuclear power plants. Nuclear negative externalities including (i) health and environmental impact, (ii) radioactive waste disposal, and (iii) the effect of severe accidents can be quantified in terms of LS units and monetary units. We were able to calculate the monetary value of NPP externalities at $277 per kilometer of distance for Kori and $280 per kilometer of distance for Wolsong at constant 2015 prices. These estimates are quite different from the traditional estimates made with the contingent valuation method, whereas they are similar to the findings of LSA studies abroad. Hence, the need to adopt the LSA in South Korea and policy implications are demonstrated.

• 한국정보통신설비학회:학술대회논문집
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• 2009.08a
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• pp.54-58
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• 2009

In this paper, the ECT(eddy current testing) signal analysis of eddy current array probe for inspection of SG(steam generator) tube in NPP(nuclear power plant) using electromagnetic FEM(finite element method) was performed. To obtain the electromagnetic characteristics of probes, the governing equation was derived from Maxwell's equation, and the problem was solved by using the 3-dimensional FEM. The types of defects were FBH(flat bottomed hole) and OD groove, Spiral groove, natural defects(pitting, SCC, multiple SCC, wear). The depth of FBH defects were 20%, 40%, 60%, 80%, 100 of SG tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequency of 100kHz, 300kHz and 400kHz were used. Material of specimen was Inconel 600 which is usually used for SG tubes in NPP. The signal difference could be observed according to the variation of size and depth on FBH defects and operation frequencies. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.502-516
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• 2002

Wolsong NPP are CANDU6 type reactors and there are 4 CANDU6 type reactors in commercial operation. CANDU type reactors require on-power refuelling by two remote controlled F/Ms (Fuelling Machines). Most of channels, fuel bundles are float by channel coolant flow and move toward downstream, however in about 30% of channels the coolant flow are not sufficient enough to carry fuel bundles to downstream. For those channels a special device, FARE (Flow Assist Ram Extension) device, is used to create additional force to push fuel bundles. It has been showing that during fuelling operation of some channels the channel coolant flow rate is reduced below specified limit (80% of normal), and consequently trip alarm signal turns on. This phenomenon occurs on selected channels that are instrumented for the channel flow and required to use the FARE device for refuelling. Hence it is believed that the FARE device causes the problem. It is also suspected that other channels that do not use the FARE device for refuelling might also go into channel flow low state. The analysis revealed that the channel How low occurs as the FARE device is introduced into the core and disappears as the FARE device is removed from the core. This paper presented the FARE device behavior, detailed fuelling operation sequence with the FARE device and effect on channel flow low phenomena. The FARE device components design changes are also suggested, such as increasing the number or now holes in the tube and flow slots in the ring orifice.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.165-169
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• 2014

In order to use the nuclear energy as the sustainable energy source, the safe and efficient management of radioactive wastes generated from the nuclear fuel cycle including NPP decommissioning is one of the most important factors. The establishment of acceptance criteria for very low level radioactive wastes generated from decommissioning of nuclear power plant in a large quantity is seemed to play a key role for developing a radioactive wastes disposal strategy as well as NPP decommissioning strategy. In this thesis, we want to review the acceptance criteria of low-and-intermediate-level radioactive wastes in this country through the analysis of other country's acceptance criteria.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• Journal of the Ergonomics Society of Korea
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• v.22 no.1
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• pp.31-42
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• 2003

This study describes a modification of the technique for human error analysis in nuclear power plants (NPPs) which adopts advanced Man-Machine Interface (MMI) features based on computerized working environment, such as LCOs. Flat Panels. Large Wall Board, and computerized procedures. Firstly, the state of the art on human error analysis methods and efforts were briefly reviewed. Human error analysis method applied to NPP design has been THERP and ASEP mainly utilizing Swain's HRA handbook, which has not been facilitated enough to put the varied characteristics of MMI into HRA process. The basic concepts on human errors and the system safety approach were revisited, and adopted the process of FMEA with the new definition of Error Segment (ESJ. A modified human error analysis process was suggested. Then, the suggested method was applied to the failure of manual pump actuation through LCD touch screen in loss of feed water event in order to verify the applicability of the proposed method in practices. The example showed that the method become more facilitated to consider the concerns of the introduction of advanced MMI devices, and to integrate human error analysis process not only into HRA/PRA but also into the MMI and interface design. Finally, the possible extensions and further efforts required to obtain the applicability of the suggested method were discussed.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.