• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1336-1342
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• 2022

Risk-informed safety classification of structures, systems and components (SSCs) is very important for ensuring the safety and economic efficiency of nuclear power plants (NPPs). However, previous methods for safety classification of SSCs do not take the plant operating modes or the operational process of SSCs into consideration, thus cannot concentrate on the safety and economic efficiency accurately. In this contribution, a new method for safety classification of SSCs based on the categorization of plant operating modes is proposed, which considers the NPPs operating history to improve the economic efficiencies while maintaining the safety. According to the time duration of plant configurations in plant operating modes, average importances of SSCs are accessed for an NPP considering the operational process, and then safety classification of SSCs is performed for plant operating modes. The correctness and effectiveness of the proposed method is demonstrated by application in an NPP's safety classification of SSCs.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.34-45
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• 2001

We have needs to develop a performance analysis system that can be used in domestic nuclear power plants to determine performance status of turbine cycle. We developed new NOPAS system to aid performance analysis of turbine cycle . Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). Robustness in the performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful aids for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis reports.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.51-55
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• 2014

Based on the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant (NPP), Common Cause Failure(CCF) events have been recognized as one of the main contributors to the risk. Also, the CCF data and estimation method used in domestic PSA models have been pointed out as an issue with respect to the quality. The existing method of MGL and non-staggered testing even widely used were considered conservative in estimating the safety and had a limited capability in uncertainty analyses. Therefore, this paper presents the CCF estimation using a new generic data source and Alpha factor method. The analyses showed that Alpha factor and staggered method are effective in estimating the CCF contribution and risk insights of reference plant. This method will be a common bases for the optimization of new design for the construction plants as well as for the updating of safety assessment on the operating nuclear power plants.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.1-6
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• 2014

PID(Proportional, Integral, Derivative) controller is one of the most used process controllers in nuclear power plants. The optimized parameter setting of process controller contributes to the stable operation and efficiency in the operating nuclear power plants. PID parameter setting is tuned when new process control system is established or process control system is changed. It is a burdensome work for I&C(Instrument and Control) engineers to tune the PID controller because it requires a lot of experience and knowledge. When the plant is in operation, inadequate PID parameter setting can be the cause of the unstable process of the plant. Therefore the results of PID parameter setting should be compared, simulated, verified and finally optimized. The practical PID tuning methods used in process controller are tuning operation calculation(Ziegler-Nicholes, Minimum TIAE, Lambda, IMC), exclusive tuning program based on computer and Matlab application. This paper introduces the various tuning methods and suggests an optimized PID tuning process in the operating nuclear power plants.

• International conference on construction engineering and project management
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• 2017.10a
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• pp.8-13
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• 2017

Nuclear power plant (NPP) constructions are large scale projects that are executed for several years, and schedule control utilizing various schedules is a critically important factor. Recently Korea independently developed the Advanced Power Reactor (APR) 1400 and is building nuclear facilities applying this new reactor type. The construction of Shin-Kori NPP (SKN) Unit 3, which adopted the APR1400, was completed and commercial operation has begun, while, SKN 4, Shin-Hanul NPP (SHN) Units 1&2, and SKN 5&6 are currently under construction. Prior to the development of the APR1400, Korea built 24 reactors and accumulated the schedule data of various reactor types which provided the foundation for schedule reduction to be possible. However, as there is no schedule development and review system established based on the standard schedule data (standard activities, durations, etc.) by reactor type, the process for developing the schedule for new builds is low in efficiency consuming much time and manpower. Also all construction data has been accumulated based on schedule activities. But because the connectivity of activities between projects is low, it is difficult to utilize such accumulated data (causes for schedule delay, causes for design changes, etc.) in new build projects. Due to such reasons, issues continue to arise in the process of developing standard schedule activities and a standard schedule for nuclear power plant construction. In order to develop a standard schedule for NPP construction, i) the development of an NPP standard schedule activity list, ii) development of the connection logic of NPP standard schedule activities, iii) development of NPP standard schedule activity resources and duration, and iv) integration of schedule data need to be performed. In this paper, an analysis was made on the coherence of schedule activity descriptions of existing NPPs by applying the probabilistic methodology on activities with low connectivity due to the utilization of the numbering system of four APR1400 reactors (SHN 1&2 and SKN 3&4).This study also describes the method for developing a standard schedule activity list and connectivity measures by extracting same and/or similar schedule activities.

• Journal of the Korea Institute of Building Construction
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• v.14 no.4
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• pp.314-321
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• 2014

Nuclear power plant projects are being increased all over the world. The construction of nuclear power plants needs huge money and time, which makes conducting a detailed analysis of productivity through the whole process. Reinforced-concrete works productivity field data was collected for more than one year and analyzed from a nuclear-power-plant project in Korea. The productivities of formwork, rebar-work, and concrete pouring were $0.54m^2/man{\cdot}day$, $0.06ton/man{\cdot}day$, $1.98m^3/man{\cdot}day$, respectively. Moreover, it is revealed that 'Day of the Week' is a driver of the formwork activity and 'Overtime' is for all of the three. The results will be a great interest of industry personnel estimating time and cost of a new nuclear power plant.

• Elastomers and Composites
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• v.55 no.3
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• pp.222-228
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• 2020

Rubber materials, used in nuclear power plants, need high heat-oxidation resistance to curing or cracking under a heat aging environment. This is because they are applied to environments with high temperature, high humidity, and radiation exposure. Nuclear radiation causes additional hardening or degradation, therefore, rubber materials need radiation resistance that satisfies the general and any accidental conditions produced in the power plant. Therefore, in this study, we developed a rubber material with excellent heat and radiation resistance for the diaphragm seal of a nuclear steam generator nozzle dam. The rubber material greatly improved the reliability of the steam generator nozzle dam. In addition, 30 inch and 42 inch diaphragm seals were manufactured using the developed rubber material. A nozzle dam was installed in a nuclear power plant and tested under the same conditions as a steam generator to evaluate safety and reliability. In the future, the performance and safety of diaphragm seals developed through field tests of nuclear power plants will be evaluated and applied to currently operating and new nuclear power plants.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.60 no.5
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• pp.1017-1022
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• 2011

This paper describes the improvement of the calculation by using $6{\sigma}$ method on steam turbine simulator in a nuclear power plant. The simulator is essential to not only verification and validation of control logic but also making sure of control constants in upgrading the long time used control system into the new one. And the dynamic model is a key point in that simulator. The model used during the retrofit period of the turbine controller in Kori Nuclear Power Plant makes difference in calculating generator output and control valve positions. That is because such operating data as the main steam pressure, the main steam temperature and control valve positions of Yongkwang #3 are different from those of Kori #4. Therefore, the model parameters must be tuned by using actual operating data for the high fidelity of simulator in calculating the dynamic characteristic of the model. This paper describes that the $6{\sigma}$ method is used in improvement of precision of generator output calculation in the steam turbine model of the simulator.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.673-682
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• 2018

One of the most widely used methods to estimate core damage during a nuclear power plant accident is containment radiation measurement. The evolution of severe accidents is extremely complex, leading to uncertainty in the containment dose rate (CDR). Therefore, it is difficult to accurately determine core damage. This study proposes to conduct uncertainty analysis of CDR for core damage assessment. First, based on source term estimation, the Monte Carlo (MC) and point-kernel integration methods were used to estimate the probability density function of the CDR under different extents of core damage in accident scenarios with late containment failure. Second, the results were verified by comparing the results of both methods. The point-kernel integration method results were more dispersed than the MC results, and the MC method was used for both quantitative and qualitative analyses. Quantitative analysis indicated a linear relationship, rather than the expected proportional relationship, between the CDR and core damage fraction. The CDR distribution obeyed a logarithmic normal distribution in accidents with a small break in containment, but not in accidents with a large break in containment. A possible application of our analysis is a real-time core damage estimation program based on the CDR.