• 한국압력기기공학회 논문집
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• 제8권2호
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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.

• 한국안전학회지
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• 제35권4호
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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.

• 한국압력기기공학회 논문집
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• 제5권1호
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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.

• 한국전산유체공학회지
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• 제20권1호
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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.

• 대한기계학회논문집A
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• 제34권9호
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• pp.1217-1225
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• 2010

소듐을 냉각재로 사용하는 고속로의 증기발생기에서는 소듐과 물의 화학적 반응을 최소화하는 것이 중요한 문제이다. 소듐과 물의 반응 가능성을 줄여 증기발생기의 신뢰성을 향상시키기 위한 한가지 방안으로 이중벽관을 전열관으로 사용하는 증기발생기를 개발하고 있다. 이 증기발생기에서 중요한 현안은 이중벽관에서의 열전달 성능을 향상시키는 문제와 원자로 운전 중에 소듐과 물 반응사고가 일어나기 전에 전열관의 파손을 감지하는 기술을 개발하는 것이다. 이 논문에서는 이 현안을 극복할 수 있는 방안을 제시하였고, 이 기술을 활용하여 증기발생기의 개념을 설계하였다. 또한 이 개념에 적용되는 이중벽관을 설계 및 예비 제작하여 기계적 시험을 수행하였다.

• 한국학교수학회논문집
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• 제16권4호
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• pp.745-770
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• 2013

오늘날 지식정보 기반 사회에서 제공되는 지식은 단일한 교과의 지식이 아니라 교과를 구분하기 힘든 통합된 형태로 나타나고 있다. 이러한 사회에서 문제해결력을 갖추기 위해서는 통합된 형태의 지식을 우선적으로 습득하고, 이를 과학적 상상이나 예술적 감성과 결합시킬 수 있는 융합적인 사고가 요구된다. 융합인재교육(STEAM)은 이러한 문제해결력과 융합적 사고를 신장시키기 위한 교육 방안의 하나로 제시되고 있다. 본 연구는 초등학교 수학과 6학년 교과서를 중심으로 수학수업에 적용할 수 있는 융합인재교육 수업자료를 개발하는 것을 목적으로 한다. 이를 위해 3단원 '각기둥과 각뿔' 수업에서는 '스파게티 프로젝트', '페이퍼 크래프트' 자료를, 4단원 '여러 가지 입체도형'에서는 'EDUCUBE' 자료를, 그리고 6단원 '비율 그래프'에서는 '나만의 팔찌 만들기' 수업자료를 개발하였다. 또한 이렇게 개발된 자료들을 실제 수업에 적용하였으며, 그 결과 특히 학생들의 수학적 태도에 있어서 긍정적인 변화를 관찰할 수 있었다. 융합인재교육을 적용한 수학수업 결과 학생들의 수업태도 및 수업에 대한 흥미가 긍정적이었으며, 수학 교과에 대한 인식이 개선된 것으로 나타났다. 이에 본 연구는 융합인재교육을 적용한 초등수학 수업자료의 개발이 보다 다양한 영역과의 융합을 통해 다양한 학년과 내용 영역에서 전개될 필요가 있음을 제안하고 있다.

• Nuclear Engineering and Technology
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• 제49권8호
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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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• 제33권4호
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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.

• 한국유체기계학회 논문집
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• 제19권4호
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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.

• 소음진동
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• 제9권2호
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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.