• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.24-32
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• 2018

System Complexity one of the most common cause failure of the projects, it leads to a lack of understanding about the functions of the system. Hence, the model is developed for communication and furthermore modeling help analysis, design, and understanding of the system. On the other hand, the text-based specification is useful and easy to develop but is difficult to visualize the physical composition, structure, and behaviour or data exchange of the system. Therefore, it is necessary to transform system description into a diagram which clearly depicts the behaviour of the system as well as the interaction between components. According to the International Atomic Energy Agency (IAEA) Safety Glossary, The safety system is a system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents. Core Protection Calculator System (CPCS) in Advanced Power Reactor 1400 (APR 1400) Nuclear Power Plant is a safety critical system. CPCS was developed using systems engineering method focusing on Departure from Nuclear Boiling Ratio (DNBR) calculation. Due to the complexity of the system, many diagrams are needed to minimize the risk of ambiguities and lack of understanding. Using Model-Based Systems Engineering (MBSE) software for modeling the DNBR algorithm were used. These diagrams then serve as the baseline of the reverse engineering process and speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.335-346
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• 2009

This paper deals with a hydroelastic vibration analysis of two rectangular plates partially coupled with a liquid, which is bounded by two plates and two rigid side walls. The wet displacement of each plate is assumed to be a combination of the modal functions of a dry uniform beam with a clamped boundary condition. As the liquid is assumed to be an ideal liquid, the displacement potential satisfying the Laplace equation is determined so that the liquid boundary conditions can meet the requirements at the rigid surfaces and the free liquid surface. The wet dynamic modal functions of each plate are expanded by using the finite Fourier transform to obtain an appropriate form of the compatibility requirement along the contacting surfaces between the plates and the liquid. The liquid-coupled natural frequencies of the plates are derived by using the Rayleigh-Ritz method. Finite element analyses using commercial software are carried out to verify the proposed theory. It is observed that the theoretical method agrees excellently with the three-dimensional finite element analyses results. The effects of the liquid depth and the liquid thickness on the normalized natural frequencies are investigated to identify the dynamic characteristics of the liquid coupled system.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1916-1938
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• 2019

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.699-714
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• 2021

Supercritical CO₂ (S-CO₂) Brayton cycle has been applied to various heat sources in recent decades, owing to the characteristics of compact structure and high efficiency. Understanding the research development in this emerging research field is crucial for future study. Thus, a bibliometric approach is employed to analyze the scientific publications of S-CO₂ cycle field from 2000 to 2019. In Scopus database, there were totally 724 publications from 1378 authors and 543 institutes, which were distributed over 55 countries. Based on the software-BibExcel, these publications were analyzed from various aspects, such as major research areas, affiliations and keyword occurrence frequency. Furthermore, parameters such as citations, hot articles were also employed to evaluate the research output of productive countries, institutes and authors. The analysis showed that each paper has been cited 13.39 times averagely. United States was identified as the leading country in S-CO₂ research followed by China and South Korea. Based on the contents of publications, existing researches on S-CO₂ are briefly reviewed from the five aspects, namely application, cycle configurations and modeling, CO₂-based mixtures, system components, and experiments. Future development is suggested to accelerate the commercialization of S-CO₂ power system.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.697-708
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• 2012

The majority of instrumentation and control (I&C) systems in today's nuclear power plants (NPPs) are based on analog technology. Thus, most existing I&C systems now face obsolescence problems. Existing NPPs have difficulty in repairing and replacing devices and boards during maintenance because manufacturers no longer produce the analog devices and boards used in the implemented I&C systems. Therefore, existing NPPs are replacing the obsolete analog I&C systems with advanced digital systems. New NPPs are also adopting digital I&C systems because the economic efficiencies and usability of the systems are higher than the analog I&C systems. Digital I&C systems are based on two technologies: a microprocessor based system in which software programs manage the required functions and a programmable logic device (PLD) based system in which programmable logic devices, such as field programmable gate arrays, manage the required functions. PLD based systems provide higher levels of performance compared with microprocessor based systems because PLD systems can process the data in parallel while microprocessor based systems process the data sequentially. In this research, a bistable trip logic in a reactor protection system (RPS) was developed using very high speed integrated circuits hardware description language (VHDL), which is a hardware description language used in electronic design to describe the behavior of the digital system. Functional verifications were also performed in order to verify that the bistable trip logic was designed correctly and satisfied the required specifications. For the functional verification, a random testing technique was adopted to generate test inputs for the bistable trip logic.

• Journal of the Korean Society of Safety
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• v.26 no.4
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• pp.87-95
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• 2011

The urgent VAI(Vital Area Identification) method development is required since 'The Act of Physical Protection and Radiological Emergency' that is established in 2003 requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The KAERI(Korea Atomic Energy Research Institute) has developed the VAI methodology and VAI software called as VIPEX(Vital area Identification Package EXpert) for identifying the vital areas. This study is to demonstrate the applicability of KAERI's VAI methodology to a hypothetical facility, and to identify the importance of information of cable and piping runs when identifying the vital areas. It is necessarily needed to consider cable and piping runs to determine the accurate and realistic TEPS(Top Event Prevention Set). If the information of cable and piping runs of a nuclear power plant is not considered when determining the TEPSs, it is absolutely impossible to acquire the complete TEPSs, and the results could be distorted by missing it. The VIPEX and FTREX(Fault Tree Reliability Evaluation eXpert) properly calculate MCSs and TEPSs using the fault tree model, and provide the most cost-effective method to save the VAI and physical protection costs.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1369-1377
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• 2021

Nowadays, it is necessary to accelerate the construction of new power plant in face of rising energy demand in such a way that the electricity will be generated at the lowest cost while reducing emissions caused by that generation. The expansion planning is one of the most important issues in electricity management. Nuclear energy comes forward with the low-carbon technology and increasing competitiveness to expand the share of generated energy by introducing Gen IV reactors. In this paper, the generation expansion planning of these new Gen reactors is investigated using the WASP software. Iran power grid is selected as a case of study. We present a comparison of the twenty-one year perspective on the future with the development of (1) traditional thermal power plants and Gen II reactors, (2) Gen III + reactors with traditional thermal power plants, (3) Gen IV reactors and traditional thermal power plants, (4) Gen III + reactors and the new generation of the thermal power plant, (5) the new generation of thermal power plants and the Gen IV reactors. The results show that the Gen IV reactors have the most developing among other types of power plants leading to reduce the operating costs and emissions. The obtained results show that the use of new Gen of combined cycle power plant and Gen IV reactors make the emissions and cost to be reduced to 16% and 72% of Gen II NPPs and traditional thermal power plants, respectively.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.3975-3984
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• 2022

Distributed control system architecture is adopted for I&C systems of Prototype Fast Breeder Reactor, where the geographically distributed control systems are connected to centralized servers & display stations via switched Ethernet networks. TCP/IP communication plays a significant role in the successful operations of this architecture. The communication tasks at control nodes are taken care by TCP/IP offload modules; local area switched network is realized using layer-2/3 switches, which are finally connected to network interfaces of centralized servers & display stations. Safety, security, reliability, and fault tolerance of control systems used for safety-related applications of nuclear power plants is ensured by indigenous design and qualification as per guidelines laid down by regulatory authorities. In the case of commercially available components, appropriate suitability analysis is required for getting the operation clearances from regulatory authorities. This paper details the proposed approach for the suitability analysis of TCP/IP communication nodes, including control systems at the field, network switches, and servers/display stations. Development of test platform using commercially available tools and diagnostics software engineered for control nodes/display stations are described. Each TCP link behavior with impaired packets and multiple traffic loads is described, followed by benchmarking of the network switch's routing characteristics and security features.

• Journal of the Korean Society of Radiology
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• v.9 no.4
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• pp.249-255
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• 2015

In this study, through the additional information of the PET / CT images by utilizing the basic data of TPS clinical application on the basis of the image re-forming synthetic function, the True-D technology and MIM software for continued research and development in combination, based on the combination-work between the respective images, reducing the time and cost of useful reading in clinical wide use of image width, efficient, effective tool for tumor targeting at diagnosis and radiation therapy by use as, by using the precise therapeutic effect determination, the time taken to read in the clinical, unnecessary and expect to a can reduce the additional examination by the creation of tumor patients read reports and PACS such asWe expect to be utilized for compatibility development with other software to evaluate the performance of PET / CT equipment.