• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• Journal of the Korean Society of Safety
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• v.35 no.4
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.25-31
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• 2009

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which comprises of the pressure boundary of primary system. The integrity of SG tube has been confirmed by the eddy current test every outage. In Korea, Bobbin probe and MRPC probe have been generally used for the eddy current test. Meanwhile the usage of Array probe has gradually increased in U.S., Japan and other countries. In this study, we investigated the defect detection capability of the Array probe through its preliminary application to SG tube inspection. The Array probe has the equivalent capability in the defect detection and sizing as the conventional methods. Thus it is desirable that the Array probe is generally applied to SG tube inspection in the domestic NPPs.

• Journal of computational fluids engineering
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• v.20 no.1
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• pp.84-91
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• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1217-1225
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• 2010

The possibility of a sodium-water reaction occurring in a conventional single-wall-tube steam generator in an SFR is a major problem. To improve the reliability of a steam generator, a double-wall-tube steam generator that can reduce the possibility of the occurrence of a sodium-water reaction is being developed. Current developments are focusing on improving the heat-transfer capability of a double-wall tube; further, the development of a leak-detection method to detect the occurrence of a sodium-water reaction during the reactor operation is also underway. In this study, new concepts, which will solve the above-mentioned problems, have been developed. Accordingly, a double-wall tube has been designed, fabricated, and mechanically tested for the purpose.

• Journal of the Korean School Mathematics Society
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• v.16 no.4
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• pp.745-770
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• 2013

In the knowledge-based society today, most knowledge is the integrated one which is difficult to be classified into subjects rather than the knowledge of a single subject. Thus, integrated thinking, which integrated knowledge is preferentially acquired first and then can be also associated with imagination and artistic sensitivity, is simultaneously required in order that we have a problem-solving capability in our daily life. STEAM education(science, technology, engineering, arts and mathematics) is one of the educational methods to improve this problem-solving capability as well as integrated thinking. This research developed materials for STEAM education which can be applied to the 6th grade curriculum of elementary school mathematics, then input it, and analyzed how it impacts with students' attitudes toward mathematics. Unit 3 'Prism' and Pyramid' were restructured and replaced by classes such as 'Spaghetti Project' or 'Paper Craft'. Unit 4 'Several Solid Figure' was taught as a class of 'EDUCUBE'. Unit 6 'Proportional Graph' was taught as a class of 'Creating my own bracelet'. After having this class, we found that mathematics class applied STEAM also has a positive effect on the mathematical attitude of students. Many students said that math is fun and gets more interesting after having math class applied STEAM and we come to know that they have positive awareness of mathematics.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1669-1679
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• 2017

Background: Steam generator (SG) is one of the significant components in the nuclear steam supply system. A variety of SGs have been designed and used in nuclear reactor systems. Every SG has advantages and disadvantages. A brief account of some of the existing SG designs is presented in this study. A high surface to volume ratio of a SG is required in small modular reactors to occupy the least space. In this paper, performance improvement for SGs of integral small modular reactor is proposed. Aims/Methods: For this purpose, cross-grooved microfins have been incorporated on the inner surface of the helical tube to enhance heat transfer. The primary objective of this work is to investigate thermal-hydraulic behavior of the proposed improvements through modeling in RELAP5-3D. Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure). The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.397-408
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• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• The KSFM Journal of Fluid Machinery
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• v.19 no.4
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• pp.29-34
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• 2016

In this study, the experimental investigation of air-cooled condensation in slightly inclined circular tubes with and without fins has been conducted. In order to assess the effects of the essential parameters, variable air velocities and steam mass flow rates were given to the test section. The heat transfer performance of air-cooled condensation were dominantly affected by the air velocity, however, the increase of the steam mass flow rate gave relatively weaker effects to total heat transfer capability. And in the experimental cases with the finned tube, the total heat transfer rate of the finned tube was significantly larger than that of the flat tube. From those results, it can be confirmed that the most important parameter for air-cooled condensation heat transfer is the convective heat transfer characteristics of air. Therefore, for the well-designed long-term cooling passive safety system, the consideration of the optimal design of the fin geometry is needed, and the experimental and numerical validations of the heat transfer capability of the finned tube would be required.

• Journal of KSNVE
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• v.9 no.2
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• pp.401-408
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• 1999

A method is presented for the vibration analysis of steam turbine-generator rotor system based on the component mode synthesis method. The motion of each component of the system is described by superposing constraint mode associated with boundary coordinates and constrained normal modes associated with internal coordinates. This method using real fixed-interface modes allows for significant reduction in system model size while retaining the essential dynamic characteristics of the lower modes. The capability of this method is demonstrated in the natural frequency and unbalance response analysis of the steam turbine-generator rotor system in which the dynamics of the pedestal is considered. The results by the present method are compared with finite element method and trnasfer matrix method in terms of the accuracy and computing time.