• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

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 high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.45-51
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• 2016

The pool water management system, which is installed for purification of the coolant in the pools and the primary cooling system of a research reactor, removes the decay heat from the reactor core when the primary cooling system stops. It also removes the heat generated from the irradiated objects in the service pool and the spent fuels in the spent fuel storage pool to keep the temperature of the pools within a limited value. In this study, the heat exchanger of the pool water management system is designed by CFD method using a commercial code Flowmaster, and the temperature of the pools is estimated along the time to conclude the design and operation method of the pool water management system.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.362-367
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• 1996

CANDU-6형 원자력발전소인 월성 2,3,4호기 비상급수계통의 성능을 평가하기 위하여 설계기능 수행과 관련된 변수로서 격납건물내 집수조(sump) 온도와 열수송계통으로 주입되는 냉각재온도를 사용한 분석을 수행하였다. 이 온도들은 NTU(Number of Transfer Unit)방법을 이용한 비상노심 냉각계통 열교환기의 열전달속도와 열전달계수의 해석을 열평형관계식과 함께 조합한 프로그램을 사용하여 계산하였다. 또한 증기발생기 급수량과 추후 수조에 공급되는 보충수에 대한 설계요건을 검토하였다. 이러한 변수와 설계요건은 비상급수계통이 발전소 정상 열제거기능 상실후 노심의 붕괴열제거에 유효한 열침원으로서의 기능을 수행함을 보여 주었다. 또 격납건물의 건전성 유지와 관련된 집수조내 최고온도가 허용치 이하로 유지되었다.

• Economic and Environmental Geology
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• v.39 no.5 s.180
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• pp.607-613
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• 2006

For the reasonable use of low grade-shallow geothermal energy by Standing Column Well(SCW) system, the basic requirements are depth-wise increase of earth temperature like $2^{\circ}C$ per every 100m depth, sufficient amount of groundwater production being about 10 to 30% of the design flow rate of GSHP with good water quality and moderate temperature, and non-collapsing of borehole wall during reinjection of circulating water into the SCW. A closed loop type-vertical ground heat exchanger(GHEX) with $100{\sim}150m$ deep can supply geothermal energy of 2 to 3 RT but a SCW with $400{\sim}500m$ deep can provide $30{\sim}40RT$ being equivalent to 10 to 15 numbers of GHEX as well requires smaller space. Being considered as an alternative of vertical GHEX, many numbers of SCW have been widely constructed in whole country without any account for site specific hydrogeologic and geothermal characteristics. When those are designed and constructed under the base of insufficient knowledges of hydrgeothermal properties of the relevant specific site as our current situations, a bad reputation will be created and it will hamper a rational utilization of geothermal energy using SCW in the near future. This paper is prepared for providing a guideline of SCW design comportable to our hydrogeothermal system.

• Proceedings of the KIEE Conference
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• 2000.11a
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• pp.262-264
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• 2000

원자로 내에서 연소 중인 핵연료나 저장 또는 재처리 중인 사용후핵연료의 성분으로서 시설의 공정설계, 안전성분석 및 차폐설계에 중요한 입력자료가 되는 핵분열생성물질, 방사화생성물 및 악티나이드의 핵종 농도와 이에 대응하는 방사능 강도의 기기 별 시간변 화율을 해석할 수 있는 코드 개발할 목적으로 MULTISAMS 정상 평형상태 모델을 구현하였다. MULTISAMS 코드의 반응공정 모델은 서로 연결되어 있으며 내부에 방사성물질의 혼합유체가 순환하는 세 종류의 반응기(원자로, 열교환기 및 화학반응기) 계통에서 자연적 또는 설계에 의해 일어나는 현상으로서; 반응기 간의 물질 흐름; 각 반응기 내에서 방사성 붕괴, 변환, 이동과 중성자 흡수 및 핵분열; 외부로부터 특정 핵종의 유입혹은 유출을 고려한 시간종속 핵종농도보존방정식 이론에 근거한다. 코드의 유용성 및 신뢰성을 검증하기 위해 현재 개념설계가 진행 중인 AMBIDEXTER원자력 에너지시스템을 대상으로 ORIGEN2 계산과 비교하였다. 두 코드 간의 입력조건과 배경이론차이점 때문에 절대적 비교가 불가능하므로 단순이론의 중간매개코드로서 SAMS를 이용한 2단계 비교방법을 따랐다. 결론은 MULTISAMS는 ORIGEN2 계산의 수렴치와 근사하게 일치하면서 ORIGEN2 가 다룰 수 없는 핵주기 연속후처리공정의 정상가동 시 핵종 평형농도를 기기 별로 계산할 수 있다는 장점을 확인하였다.