• Proceedings of the KAIS Fall Conference
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• 2008.11a
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• pp.365-367
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• 2008

원자로 냉각계통의 압력경계를 구성하고 있는 재료들의 부식은 재료 표면에 형성되는 산화막, 금속재료의 구성성분이 용해되어 생성된 가용성 화학종 및 산화물 입자 형태의 부식생성물들을 발생시킨다. 금속합금의 부식에 의한 가용성 화학종 및 입자들의 방출은 원자로 냉각계통에서 노심과 증기발생기를 순환하면서 연료피복관 위에 침전되어 여러 가지 문제를 야기한다. 크러드는 구조재료의 부식에 기인하여 발생한 부식생성물들이 냉각수에 부유하여 떠다니거나 피복관 표면에 침적하여 형성되며 주로 니켈과 철 산화물로 구성되어 있다. 원자로 냉각계통에서 크러드를 최소화하기 위하여 수화학 조건들을 제어하지만 장주기 고연소도 노심에서 AOA 현상을 일으키는 주된 원인이 되고 있다. 피복관 위에 침적되는 크러드는 붕소의 잠복위치를 제공할 뿐만 아니라 냉각수의 압력강하를 증가시키고 피복관의 부식 및 파손 원인을 제공하며 방사선 준위가 증가하도록 한다. 따라서 본 연구에서는 반응속도론적 관점에서 원자로 정지시의 용출 크러드 특성에 대한 연구를 수행하였다.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.444-447
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• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor (PWR) or a heavy water power reactor (CANDU). There is an in-pile section (IPS) and an out-pile section (OPS) in this test loop. When HANARO is normally operated, the fuel loaded in the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. The pump can not continuously suck a fluid and not pressurize the fluid during a cold function test. To verify the flow characteristics of the MCWS, a flow net work analysis has been conducted. When the higher elevation pipelines wholly filled with coolant, it was confirmed through the analysis results that the pump pressurized the coolant normally. And the analysis results described the system characteristics with operation temperature and pressure variation satisfactorily.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.67-72
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• 1996

수격현상(Waterhammer)으로 인한 과도압력하중은 월성원자력발전소 비상노심냉각계통 (Emergency Core Cooling System : ECCS) 설계의 주요 고려사항이다. 비상노심냉각계통은 특수안전계통으로서 냉각재상실사고(Loss of Coolant Accident : LOCA)후 일차열수송계통을 다시 채워주고 핵연료 손상을 막기위해 노심으로부터 잔열 및 붕괴열을 제거한다. 일차열수송계통으로의 비상냉각수 주입은 고압주입, 중압주입, 저압주입 3 단계로 주입된다. 과도압력이 발생될 것으로 예상되는 고압주입과 중압주입에 대한 6가지 사례들이 ECCS의 배관과 지지대 설계를 위해 고려되었다. 모든 사례에 대한 비상노심냉각계통의 과도압력 현상은 PTRAN 코드에 의해 해석 되었고 해석된 최고과도압력은 설계압력보다 작음을 알게 되었다. 모든 사례의 최고압력과 최고차압은 비상노심냉각계통 배관 및 지지대 설계를 위한 응력해석 자료로서 사용될 것이다.

• Journal of Energy Engineering
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• v.28 no.4
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• pp.103-110
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• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.

• Solar Energy
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• v.15 no.3
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• pp.59-73
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• 1995

The thermal analysis is accomplished with the route for the underground power transmission system which adopts the water pipe cooling with trough in tunnel. As a result, in case of a cooling system based on a refrigerator, the optimum condition for the flow rate of cooling water and the air velocity are calculated as the $2{\sim}3{\ell}/s/pipe$ and $1{\sim}2m/s/fan$, respectively. On the other hand, in case of cooling tower the optimum condition for them are calculated as the $2{\sim}3{\ell}/s/pipe$ and 6 m/s/fan, respectively. But the cooling system based on a cooling tower has the problem of enlarging the size of cooling fan and suppressing the labor of operator in tunnel. Therefore, to meet all the cooling conditions for a given cooling section, the cooling system based on a refrigerator is more acceptable.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.525-530
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• 1995

최적 열수력 전산 코드인 CATHARE2 Vl.3u 코드를 이용하여 영광 3/4호기 midloop 운전중 잔열제거(RHR) 기능 상실사고를 해석하였다. 본 연구의 주된 목적은 사고시 계통에서 발생하는 열수력 현상의 이해 향상 및 증기발생기 열제거 능력 평가에 있다. 사고 복구 절차 관점에서 노심 비등, 노출 시점 및 계통압력 등이 중요한 인자이다. 본 계산 수행시 사용한 가정은 다음과 같다. 가) 초기 계통 수위는 고온관 중간에 위치하며 그 윗 부분은 질소 가스로 차 있다. 나) 3/4 인치 크기의 방출 밸브가 원자로 용기 상부 및 가압기 상부에 각각 설치되어 있으며, RHR 흡입구에 수위지시계가 설치되어 있다. 다) 증기발생기의 이차측은 U-튜브가 잠기도록 물로 차있다. 라) 두 증기발생기의 대기 방출 밸브(ADV)는 항상 열려 있어 사고시 이차측 압력을 대기압으로 유지하기에 충분하다. 사고는 원자로 정지 2일 후 발생하였다고 가정한다. 이와 같은 조건하에서 사고시 주된 계통 열제거 수단은 증기발생기 U-튜브내의 응축 작용이며 이는 전체 열제거량의 94%로 나타났다. 노심 비등 시점온 사고후∼300초 이후이며, 계통압력은 10,800초 이후에 최고 압력인 0.25MPa에 도달한 후 그 값을 계속 유지하고 있다. RHR 배관에 연결된 수위지 시계를 통해 10,200초 이후부터 냉각수가 방출되었다. 2개의 방출밸브 및 수위지시계를 통하여 방출된 유량에 근거하여 원자로 용기 냉각재 수위가 고온관 바닦까지 낮아지는 시점을 계산하면 사고 약 6.4 시간 이후가 된다.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2505-2510
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• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. When HANARO is normally operated, the fuel loaded in the irradiation hole has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. This paper describes the cold function test results of the MCWS. It was confirmed through the test results that the system met the design requirements under a cold operation condition.

• Journal of Energy Engineering
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• v.8 no.4
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• pp.525-532
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• 1999

The standardized Korea Next Generation Reactor (KNGR) NSSS has developed in the basis of the ABB-CE System 80+ design concept. In this study, several regulatory requirements for the KNGR shutdown cooling system (SCS) operation are investigated. The purpose of this study is to establish the technical self-reliance for SCS design by supporting fundamental data such as SDCHX effective area and reactor CCW flow rate. Thermal power of KNGR would be increased to about 4,000 $MW_{th}$ in comparison with thermal power 2.825 $MW_{th}$ of UCN 3&4, therefore, SCS design data shall b recalculated by using the KDESCENT Code, which could evaluate cooling capability of SCS. It is found that SCS minimum flow rate is able to remove the primary sensible heat. Reviewing the major components such as heat exchanger, pump, value, and operating procedure, it is concluded as follows.

• Korean Journal of Poultry Science
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• v.38 no.2
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• pp.97-104
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• 2011

We conducted a 10-month (March to October 2009) surveillance of Newcastle disease virus (NDV) in 13 slaughterhouses in Korea. NDV was isolated in 13.0%, 13.3%, 16.0%, and 10.8% of chicken farms, transport vehicles, hang rooms, and chilling water, respectively. Of NDV isolates from slaughterhouses, 37% were isolated in July. All NDV isolates were determined to be lentogenic viruses by RT-PCR-based pathotyping, and all NDV isolates had the $^{112}GKQGR/L^{117}$ motif at the cleavage site of the F protein. Phylogenetic analysis based on the hypervariable region of the F protein gene classified all 25 NDV isolates examined into genotype I within class II. Of these, 24 were clustered together with the NDV V4 strain, while the remaining isolate was placed in the cluster belonging to the NDV Ulster 2C strain. Our results indicate that lentogenic NDV was a high-frequency contaminant in the serial process of ranging live birds to slaughtering at slaughterhouses.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.