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.

KALIMER 고온풀 자유액면 거동 해석

김성오(Seong-O Kim),어재혁(Jae-Hyuk Eoh),최훈기(Hoon-Ki Choi) 한국전산유체공학회 2002 한국전산유체공학회지 Vol.7 No.3

An analytic methodology was developed for free surface motions between liquid metal coolant and cover gas in order to calculate the phenomena of gas entrainment in hot pool surface through IHX EMF and reactor core. The methodology was setup by applying the first order VOF convection model to CFX4 general purpose fluid dynamics analysis code. The methodology was validated by applying it to an experimental apparatus designed for free surface motions of KALIMER reactor. The distributions of free surface calculated by the present methodology were almost coincident with the experimental data. The developed methodology was applied to the KALIMER reactor of full power operating condition. The shapes of the free surface were nearly uniform. From the results, it was found that the altitude of the free surface from the IHX inlet nozzle of KALIMER reactor is high enough not to affect to free surface motions of generating gas bubbles from the turbulent shear flows such as hydraulic jump and water falls.

소듐냉각고속로 붕괴열교환기의 고온 설계 및 건전성 평가

이형연(Hyeong-Yeon Lee),어재혁(Jae-Hyuk Eoh) 대한기계학회 2013 大韓機械學會論文集A Vol.37 No.10

본 연구에서는 소듐냉각 고속로 붕괴열교환기(DHX)의 고온 설계 및 크리프-피로 손상 평가를 수행하였다. 제 4 세대 소듐냉각 고속로의 능동 및 피동 잔열제거계통에 설치되는 DHX 와 한국원자력연구원의 STELLA-1 시험루프에 설치된 DHX 에 대해 상세설계 및 3D 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준인 ASME Section Ⅲ Subsection NH 와 RCC-MR 코드를 따라 크리프-피로 손상평가를 수행하였다. 크리프-피로 손상평가 결과에 기초하여 두 설계기준에 대해 비교 분석하고, 설계 기술기준의 보수성 이슈에 대해 토의하였다. In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section Ⅲ Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.

소듐 시험루프 내 소듐대 공기 열교환기의 고온 설계

이형연(Hyeong-Yeon Lee),어재혁(Jae-Hyuk Eoh),이용범(Yong-Bum Lee) 대한기계학회 2013 大韓機械學會論文集A Vol.37 No.5

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다. In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the hightemperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

중간 열교환기 높이 상승에 의한 KALIMER-600 원자로 풀 과도 성능 변화 분석

한지웅(Ji-Woong Han),어재혁(Jae-Hyuk Eoh),김성오(Seong-O Kim) 대한기계학회 2010 大韓機械學會論文集B Vol.34 No.11

소듐냉각 고속로 내부기기 배치 변경에 의한 초기냉각 성능변화를 검토하기 위하여 중간열교환기의 수직배치가 다른 3개의 원자로를 대상으로 COMMIX-1AR/P 코드를 활용한 다차원 해석을 수행하였다. 원통좌표계의 중심축을 기준으로 원주방향의 1/4 부분만을 모델링하고 정상상태 및 과도상태 분석을 수행하여 IHX 수직배치 변화가 초기 냉각 특성에 미치는 영향을 분석하였고, DHX를 통한 후기 냉각 모드 개시 시점에 미치는 영향도 분석하였다. 분석 결과 IHX 수직배치 상승은 원자로 풀내부 자연순환 유량을 증가시켜 초기 냉각과정에서 노심 최고 온도의 급격한 상승을 방지할 수 있으며, 초기냉각성능을 향상시키기 위한 관성회전차의 가용설계재원의 범위도 확대시킨다. 또한 IHX 수직배치 상승은 후기냉각모드에 큰 영향을 주지 않으면서 초기냉각성능의 향상에 기여할 수 있을 것으로 사료된다. The effect of increasing the elevation of an IHX (intermediate heat exchanger) on the transient performance of the KALIMER-600 reactor pool during the early phase of a loss of normal heat sink accident was investigated. Three reactors equipped with IHXs that were elevated to different heights were designed, and the thermal-hydraulic analyses were carried out for the steady and transient state by using the COMMIX-1AR/P code. In order to analyze the effects of the elevation of an IHX between reactors, various thermal-hydraulic properties such as mass flow rate, core peak temperature, RmfQ (ratio of mass flow over Q) and initiation time of decay heat removal via DHX (decay heat exchanger) were evaluated. It was found that with an increase in the IHX elevation, the circulation flow rate increases and a steep rise in the core peak temperature under the same coastdown flow condition is prevented without a delay in the initiation of the second stage of cooling. The available coastdown flow range in the reactor could be increased by increasing the elevation of the IHX.

펌프 회전차의 관성모멘트 제공에 의한 KALIMER-600 원자로 풀 과도 성능 분석

한지웅(Ji-Woong Han),어재혁(Jae-Hyuk Eoh),이태호(Tea-Ho Lee),김성오(Seong-O Kim) 대한기계학회 2009 大韓機械學會論文集B Vol.33 No.6

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.

소듐냉각 실증로 공기열교환기의 고온 설계 및 손상평가

이형연(Hyeong-Yeon Lee),어재혁(Jae-Hyuk Eoh),김종범(Jong-Bum Kim),박홍윤(Hong-Yune Park) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.5-2

In a Korean Generation IV demonstration sodium-cooled fast reactor (SFR), various types of high temperature heat exchangers such as IHX (Intermediate heat exchanger), DHX (Decay heat exchanger), AHX (Air heat exchanger) and SG (Steam Generator) are to be designed and installed. In this study, high temperature conceptual design and damage evaluation for the sir-to-sodium heat exchanger, AHX operating at creep regime in a 600MWe Demonstration Sodium-cooled Fast Reactor (SFR) have been performed according to recent version of the high temperature design codes of the ASME Section Ⅲ Subsection NH and RCC-MR. Evaluations of creep-fatigue damage based on full 3D finite element analysis were conducted according to the high temperature design codes and code comparisons were made.

소듐 시험루프 내 소듐대 공기 열교환기의 고온 설계

이형연(Hyeong-Yeon Lee),어재혁(Jae-Hyuk Eoh),이용범 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.11

In a Korean Generation Ⅳ prototype sodium-cooled fast reactor (SFR), various types of high temperature heat exchangers such as IHX (Intermediate heat exchanger), DHX (Decay heat exchanger), AHX (Air heat exchanger), FHX (Finned-tube sodium-to-Air heat exchanger) and SG (Steam Generator) are to be designed and installed. In this study, high temperature basic design of the FHX in SEFLA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) test loop to be installed at KAERI(Korea Atomic Energy Research Institute) and AHX in STELLA-1(Sodium integral effect test loop for safety simulation and assessment) test loops already installed at KAERI in a prototype SFR have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted according to the high temperature design codes and code comparisons were made for AHX.

Design and analysis of high-temperature piping system in the STELLA-2 sodium test facility

Hyeong-Yeon Lee(이형연),Jae-Hyuk Eoh(어재혁),Seok-Kwon Son(손석권),Jong-Bum Kim(김종범),Ji-Young Jeong(정지영),Yong-Sun Ju(주용선) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.2015 No.11

Elevated temperature design and integrity evaluation have been conducted for the high-temperature piping systems in the STELLA(Sodium integral effect test loop for safety simulation and assessment)-2 sodium test facility which is under development at KAERI(Korea Atomic Energy Research Institute). There are several piping systems in the STELLA-2 sodium test loop. Piping systems were designed and design evaluation were conducted for the piping systems in IHTS(Intermediate Heat Transport System) and PSLS(Pump Simulation Loop System). Piping analyses for these sodium test facility have been performed according to the piping design codes of ASME B31.1 and integrity of the piping systems were confirmed.

핀형 소듐대 공기 열교환기의 고온 설계 및 건전성 평가

이형연(Hyeong-Yeon Lee),어재혁(Jae-Hyuk Eoh),이제환(Je-Whan Lee),이용범(Yong-Bum Lee) 대한기계학회 2013 대한기계학회 춘추학술대회 Vol.2013 No.3