• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2014.10a
• /
• pp.102-105
• /
• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• International conference on construction engineering and project management
• /
• 2017.10a
• /
• pp.8-13
• /
• 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 computational fluids engineering
• /
• v.10 no.3 s.30
• /
• pp.9-18
• /
• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Journal of the Korean Society of Systems Engineering
• /
• v.13 no.1
• /
• pp.25-31
• /
• 2017

This paper aims to evaluate the conservatism in the design of APR1400 (Advanced Pressurized water Reactor 1400 designed by KHNP) reactor internals component, the LSS (Lower Support Structure). Re-engineering of the LSS is done based on the system design condition data and applicable ASME code that was used for the original APR1400 design. Systems engineering approach is applied to design the LSS of APR1400 without refering APR1400 LSS dimensional parameters and tries to verify important design parameters of APR1400 LSS as well as the validity of the re-engineering design process as independent verification method of reactor component design. Systems engineering approach applied in this study following V-model approach. The re-engineered LSS design showed more than enough conservatism for static loading case. The maximum deflection of LSS is under 1mm (calculated value is 0.25mm) from 4000 mm diameter of LSS. Hence the deflection can be ignored in other reactor internals for structural integrity assessment. Especially the effect of LSS deflection on fuel assembly can be minimized and which is one of the main requirements of LSS design. It also showed that the maximum stress intensity is 2.36MPa for the allowable stress intensity of 60.1 MPa. The stress resulted from the static load is also very small compared to the maximum allowable stress intensity, hence there is more than enough conservatism in the LSS design.

• Journal of the Korean Society of Systems Engineering
• /
• v.12 no.1
• /
• pp.73-87
• /
• 2016

Combining power generation and water production by desalination is economically advantageous. Most desalination projects use fossil fuels as an energy source, and thus contribute to increased levels of greenhouse gases. Environmental concerns have spurred researchers to find new sources of energy for desalination plants. The coupling of nuclear power production with desalination is one of the best options to achieve growth with lower environmental impact. In this paper, we will per-form a sensitivity study of coupling nuclear power to various combinations of desalination technology: {1} thermal (MSF [Multi-Stage Flashing], MED [Multi-Effect Distillation], and MED-TVC [Multi-Effect Distillation with Thermal Vapour Compression]); {2} membrane RO [Reverse Osmosis]; and {3} hybrid (MSF-RO [Multi-Stage Flashing & Reverse Osmosis] and MED-RO [Multi-Effect Distillation & Reverse Osmosis]). The Korean designed reactor plant, the APR1400 will be modeled as the energy production facility. The economical evaluation will then be executed using the computer program DEEP (Desalination Economic Evaluation Program) as developed by the IAEA. The program has capabilities to model several types of nuclear and fossil power plants, nuclear and fossil heat sources, and thermal distillation and membrane desalination technologies. The output of DEEP includes levelized water and power costs, breakdowns of cost components, energy consumption, and net saleable power for any selected option. In this study, we will examine the APR1400 coupled with a desalination power plant in the Kingdom of Saudi Arabia (KSA) as a prototypical example. The KSA currently has approximately 20% of the installed worldwide capacity for seawater desalination. Utilities such as power and water are constructed and run by the government. Per state practice, economic evaluation for these utilities do not consider or apply interest or carrying cost. Therefore, in this paper the evaluation results will be based on two scenarios. The first one assumes the water utility is under direct government control and in this case the interest and discount rate will be set to zero. The second scenario will assume that the water utility is controlled by a private enterprise and in this case we will consider different values of interest and discount rates (4%, 8%, & 12%).

• Journal of the Korean Society of Systems Engineering
• /
• v.13 no.2
• /
• pp.9-17
• /
• 2017

Concept development of a real-time Field Programmable Gate Array (FPGA)-based switched Ethernet data communication network for the Man-Machine Interface System (MMIS) is presented in this paper. The proposed design discussed in this research is based on the systems engineering (SE) approach. The design methodology is effectively developed by defining the concept development stage of the life-cycle model consisting of three successive phases, which are developed and discussed: needs analysis; concept exploration; and concept definition. This life-cycle model is used to develop an FPGA-based time-triggered Ethernet (TTE) switched data communication network for the non-safety division of MMIS system to provide real-time data transfer from the safety control systems to the non-safety division of MMIS and between the non-safety systems including control, monitoring, and information display systems. The original IEEE standard 802.3 Ethernet networks were not typically designed or implemented for providing real-time data transmission, however implementing a network that provides both real-time and on-demand data transmission is achievable using the real-time Ethernet technology. To develop the design effectively, context diagrams are implied. Conformance to the stakeholders needs, system requirements, and relevant codes and standards together with utilizing the TTE technology are used to analyze, synthesize, and develop the MMIS non-safety data communication network of the APR1400 nuclear power plant.

• Nuclear Engineering and Technology
• /
• v.36 no.1
• /
• pp.24-35
• /
• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.64 no.2
• /
• pp.254-259
• /
• 2015

On January 30, 2012 an auxiliary component of Byron Unit 2 was tripped on bus under voltage. The cause of the event was the failure of the C-phase insulator track for the Unit 2 station auxiliary transformer(SAT) revenue metering transformer. In addition to this event, other events have occurred at other plants resulting in an open-phase condition.[1] Therefore, Nuclear Regulatory Commission(NRC) has requested that not only nuclear power plant(NPP) operating company but also its Design Certification(DC) applicant have to prepare open-phase detecting system in their operating plants and design document. In this paper, various open-phase conditions are simulated in NPP using Electromagnetic Transient Program(EMTP) and Atpdraw, and open-phase condition detecting system is proposed for Main Transformer(MT), Unit Auxiliary Transformer(UAT) and SAT connected power line in NPP.

• Journal of the Korean Society of Systems Engineering
• /
• v.15 no.1
• /
• pp.51-59
• /
• 2019

After Fukushima Dai-ichi NPP accident, the need for implementation of diverse and flexible coping strategies (FLEX) became evident. However, to ensure the effectiveness of the safety strategy, it is essential to quantify the uncertainties associated with the station blackout (SBO) scenario as well as the operator actions. In this paper, a systems engineering approach for uncertainty analysis (UA) of a SBO scenario in advanced pressurized water reactor is performed. MARS-KS is used as a best estimate thermal-hydraulic code and is loosely-coupled with Dakota software which is employed to develop the uncertainty quantification framework. Furthermore, the systems engineering approach is adopted to identify the requirements, functions and physical architecture, and to develop the verification and validation plan. For the preliminary analysis, 13 uncertainty parameters are propagated through the model to evaluate the stability and convergence of the framework. The developed framework will ultimately be used to quantify the aleatory and epistemic uncertainties associated with an extended SBO accident scenario and assess the coping capability of APR1400 and the effectiveness of the implemented FLEX strategies.

• 한국전산유체공학회:학술대회논문집
• /
• 2009.11a
• /
• pp.121-128
• /
• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.