나트륨-물 반응 사고 중/후반기 거동특성 연구

정지영,어재혁,정경채,김병호,김종만,김태준,박남국 한국공업화학회 2003 응용화학 Vol.7 No.1

Generally, the analysis of a sodium-water reaction (SWR) is classified by two major events. One is the very initial stage of peak pressure and shock wave propagation caused by the reaction itself, and the other is the bulk motion including the mass transfer phase in the quasi-steady state of the reaction period during several second or minute orders after leak initiation. The system dynamic response during a sodium-water reaction event shows very different characteristics between the initial stage and the long-term period. In order to investigate the later phase of a SWR event. the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the desing and construction of the test facility has been made.

소듐-물 반응사고 중/후반기 계통거동특성 연구

정지영,어재혁,김태준,정경체,김병호,한도희,박남국 한국공업화학회 2004 응용화학 Vol.8 No.1

In order to investigate the later phase of a sodium-water reaction (SWR) event in KALIMER (Korea Advanced Liquid MEtal Reactor), the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and an experimeantal study has been carried out for verification of the simple analysis model applied to the code. The 24 data set obtained in the experiment have been pre-analyzed. Comparison of SELPSTA results with experimental data shows fairly good agreement in the transient.

이중벽튜브 Na-CO₂ 열교환기의 열적특성 연구

위명환,김성오,어재혁,김진환 한국공업화학회 2005 응용화학 Vol.9 No.1

A Na-CO₂ heat exchanger is one of the key equipments in the concept of coupling the supercritical CO₂ Brayton cycle to the KALIMER-150 design. The exchanger is a shell and tube counter flow type with a hot sodium on the shell side and a supercritical carbon dioxide on tube side. In this study, a computational design methodology was developed and the sensitivity study was done on the major design parameters for the development of preliminary design. In the designing process, the design requirements are pressure drops and sodium maximum velocity in the shell side and aspect ratio of the heat exchanger. The evaluation revealed that the gap design of doubel wall tube has noticeable effects on the thermal-hydraulic characteristics of the heat changer. Maintaining the operating condition to be identical to those of KALIMER-150, the heat transfer area of He packed case was 4 times more than that of LBE packed case.

반달형 베플을 갖는 이중벽튜브 중기발생기의 특성 평가

위명환,김성오,전원대,어재혁,김진환 한국공업화학회 2004 응용화학 Vol.8 No.2

The DWTSG concept is proposed as a advanced steam generator of liquid metal reactor to avoid sodium-water reaction incidents in the SG. The optimization of steam generator from economic viewpoint is to minimize the total cost of heat transfer area and pumping power. The baffles are primarily used in shell and tube steam generator for inducing cross flow over tube bundles, and as a results, improving heat transfer performance. In this paper a computer based design method, which covers segmentally baffled steam generator is developed for preliminary DWTSG design. And parametric sensitivity studies are done to determine required heat transfer area to meet the specified heat capacity by calculating minimum or allowable shell-side pressure loss.

Wastage and Self-Plugging by a Potential CO₂ Ingress in a Supercritical CO₂ Power Conversion System of an SFR

EOH, Jae-Hyuk,NO, Hee Cheon,YOO, Yong-Hwan,JEONG, Ji-Young,KIM, Jong-Man,KIM, Seong-O Atomic Energy Society of Japan 2010 Journal of nuclear science and technology Vol.47 No.11

<P>For a CO<SUB>2</SUB> ingress accident into liquid sodium in a supercritical CO<SUB>2</SUB> power conversion system coupled with a sodium-cooled fast reactor, we investigated two major design issues: i) a wastage phenomenon in regard to structural damage adjacent to the leaking position, and ii) potential channel plugging due to the formation of a particulate reaction product. In order to understand the factors affecting the occurrence of these issues, two kinds of experiments were carried out: a wastage effect test and a self-plugging test. All experimental conditions were chosen to reasonably represent the normal operating conditions and realistic design parameters of the reference plant. The test results indicate the absence of wastage, which will not lead to additional tube ruptures and damage propagation. In the current experiment, the self-plugging of PCHE channels only took place under two limited conditions: i) the sodium temperature is over 500°C and ii) the equivalent diameter of the crack opening is less than 1.5 mm with a small leakage rate of far less than 1 g/s of CO<SUB>2</SUB> ingress.</P>

Effects of Condensation Heat Transfer Model in Calculation for KNGR Containment Pressure and Temperature Response

Eoh, Jae-Hyuk,Park, Shane,Jeun, Gyoo-Dong,Kim, Moo-Hwan Korean Nuclear Society 2001 Nuclear Engineering and Technology Vol.33 No.2

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

DEVELOPMENT OF A SUPERCRITICAL CO₂ BRAYTON ENERGY CONVERSION SYSTEM COUPLED WITH A SODIUM COOLED FAST REACTOR

JAE-EUN CHA,TAE-HO LEE,JAE-HYUK EOH,SUNG-HWAN SEONG,SEONG-O KIM,DONG-EOK KIM,MOOHWAN KIM,TAE-WOO KIM,KYUN-YUL SUH 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.8

Systematic research has been conducted by KAERI to develop a supercritical carbon dioxide Brayton cycle energy conversion system coupled with a sodium cooled fast reactor. For the development of the supercritical CO₂ Brayton cycle ECS, KAERI researched four major fields, separately. For the system development, computer codes were developed to design and analyze the supercritical CO₂ Brayton cycle ECS coupled with the KALIMER-600. Computer codes were developed to design and analyze the performance of the major components such as the turbomachinery and the high compactness PCHE heat exchanger. Three dimensional flow analysis was conducted to evaluate their performance. A new configuration for a PCHE heat exchanger was developed by using flow analysis, which showed a very small pressure loss compared with a previous PCHE while maintaining its heat transfer rate. Transient characteristics for the supercritical CO₂ Brayton cycle coupled with KALIMER-600 were also analyzed using the developed computer codes. A Na-CO₂ pressure boundary failure accident was analyzed with a computer code that included a developed model for the Na-CO₂ chemical reaction phenomena. The MMS-LMR code was developed to analyze the system transient and control logic. On the basis of the code, the system behavior was analyzed when a turbine load was changed. This paper contains the current research overview of the supercritical CO₂ Brayton cycle coupled to the KALIMER-600 as an alternative energy conversion system.

DEVELOPMENT OF A SUPERCRITICAL CO₂ BRAYTON ENERGY CONVERSION SYSTEM COUPLED WITH A SODIUM COOLED FAST REACTOR

Cha, Jae-Eun,Lee, Tae-Ho,Eoh, Jae-Hyuk,Seong, Sung-Hwan,Kim, Seong-O,Kim, Dong-Eok,Kim, Moo-Hwan,Kim, Tae-Woo,Suh, Kyun-Yul Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.8

Systematic research has been conducted by KAERI to develop a supercritical carbon dioxide Brayton cycle energy conversion system coupled with a sodium cooled fast reactor. For the development of the supercritical $CO_2$ Brayton cycle ECS, KAERI researched four major fields, separately. For the system development, computer codes were developed to design and analyze the supercritical $CO_2$ Brayton cycle ECS coupled with the KALIMER-600. Computer codes were developed to design and analyze the performance of the major components such as the turbomachinery and the high compactness PCHE heat exchanger. Three dimensional flow analysis was conducted to evaluate their performance. A new configuration for a PCHE heat exchanger was developed by using flow analysis, which showed a very small pressure loss compared with a previous PCHE while maintaining its heat transfer rate. Transient characteristics for the supercritical $CO_2$ Brayton cycle coupled with KALIMER-600 were also analyzed using the developed computer codes. A Na-$CO_2$ pressure boundary failure accident was analyzed with a computer code that included a developed model for the Na-$CO_2$ chemical reaction phenomena. The MMS-LMR code was developed to analyze the system transient and control logic. On the basis of the code, the system behavior was analyzed when a turbine load was changed. This paper contains the current research overview of the supercritical $CO_2$ Brayton cycle coupled to the KALIMER-600 as an alternative energy conversion system.

초임계 이산화탄소 Brayton 에너지 전환계통 예비설계

차재은(Jae-Eun Cha),어재혁(Jae-Hyuk Eoh),이태호(Tae-Ho Lee),성승환(Sung-Hwan Sung),김성오(Seong-O Kim),김태우(Tae-Woo Kim),김동억(Dong-Eok Kim),김무환(Moo-Hwan Kim) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.11

The supercritical CO₂ Brayton cycle energy conversion system is presented as a promising alternative to the present Rankine cycle. The principal advantage of the S-CO₂ gas is a good efficiency at a modest temperature and a compact size of its components. The S-CO₂ Brayton cycle coupled to a SFR also excludes the possibilities of a SWR (Sodium-Water Reaction) which is a major safety-related event, so that the safety of a SFR can be improved. KAERI is conducting a feasibility study for the supercritical carbon dioxide (S-CO₂) Brayton cycle power conversion system coupled to KALIMER(Korea Advanced LIquid MEtal Reactor). The purpose of this research is to develop S-CO₂ Brayton cycle energy conversion systems and evaluate their performance when they are coupled to advanced nuclear reactor concepts of the type under investigation in the Generation Ⅳ Nuclear Energy Systems. This paper contains the research overview of the S-CO₂ Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system.

High temperature design and damage evaluation of a helical type sodium-to-air heat exchanger in a sodium-cooled fast reactor

이형연,Jae-Hyuk Eoh,이용범 대한기계학회 2013 JOURNAL OF MECHANICAL SCIENCE AND TECHNOLOGY Vol.27 No.9

A high temperature design and creep-fatigue damage evaluation for a helical type sodium-to-air heat exchanger (AHX) in a 600MWe sodium-cooled fast reactor (SFR) has been conducted. The AHX is a safety-grade component in a passive decay heat removal (DHR)system. Safe and reliable decay heat removal is one of the most important tasks in the successful design of a sodium-cooled fast reactor. Therefore, independent, diverse and redundant DHR systems incorporating both passive and active mechanisms have been employed in the demonstration SFR developed by Korea Atomic Energy Research Institute (KAERI). One of the key DHR components is sodium-toair heat exchanger, which has a helically coiled tube arrangements. A creep-fatigue damage evaluation has been performed according to the elevated temperature design codes of ASME Subsection NH and RCC-MRx based on a full 3D finite element analysis. The integrity of the heat exchangers under creep-fatigue loading was confirmed and code comparisons were made as well.