원자력발전소 분기관 내의 열성층 실험 연구

김상녕,황선홍 경희대학교 산학협력기술연구원 2004 산학협력기술연구논문집 Vol.10 No.2

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line , steam generator inlet nozzle, safety injection system(SIS), and chemical and volume control system(CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping defacement, and pipe support damage. The phenomena is one of the unaccounted load in the design stage. However, the load found to be serious as nuclear power plant operation experience accumulated. In particular, the thermal stratification by the turbulent penetration or value leak in the SIS and SCS pipe line can lead the system failure by the thermal fatigue. Therefore in this study an experimental work had been executed to predict temperature distribution (thermal stratification) in these system by the turbulent peroration and valve leak. Also an analytical model was developed to predict the amount of heat transfer through wall and the valve leak. The results are expected to be useful to understand the thermal stratification caution in the system and validate the code calculation results by such as Fluent, Phenix, CFX

단상류와 이상류에서 펌프특성간의 상호관계 결정

김상녕,김순기,안성수 慶熙大學校 材料科學技術硏究所 1989 材料科學技術硏究論集 Vol.2 No.-

Using a 1/10 scale model pump, designed and manufactured to simulate two phase performance of reactor coolant pump of Y.G.N #3 and 4, a set of experiments was executed with water and air of cold states (at atmosphere and room temperature). A head loss coefficient, which is defined as a nondimensional ratio of the differnece of theoretical and actual pump head coefficient in single phase to that in two phase, , H^(*) = (Ψ_(th) - Ψ_(act)) t·p / (Ψ_(th) - Ψ_(act)) s·p , was determined as a function of void fraction (α) and flow coefficient (Ø). The calibrated mass flow rates of air and water were used to calculate the void fraction by using drift flux model. In particular, flow fluctuation starts when the void fraction is 0.02, then as the void fraction increases it becomes more severe, and finally flow stops when the void fraction is 0.10.

IRWST의 증기응축 및 열성층 현상 연구

김상녕,고병만,윤기훈 경희대학교 산학협력기술연구원 2004 산학협력기술연구논문집 Vol.10 No.2

The In-Containment Refueling Water Storage Tank (IRWST), one of the design improvements applied to the APR-1400, has a function to condense the high enthalpy fluid discharged from the Reactor Coolant System (RCS). The condensation of discharged fluid by the tank water drives the tank temperature high and causes oscillatory condensation. Also if the tank cooling water temperature approaches the saturated state, the steam bubble may escape from the water uncondensed. These oscillatory condensation and bubble escape would burden the undue load to the tank structure, pressurize the tank, and degrade its intended function. For these reasons experimental works was performed in order to predict exact tank temperature distribution and to find the effective cooling method to keep the tank temperature below the bubble escape limit (93.3℃), which was experimentally proven by other researchers. The diameter of the experimental system was scaled down to 1:20compared with IRWST of APR-1400, and the temperature distribution was measured by thermocouple. Also the steam, which needed on experiment, was made by Boiler which had 1000kg/hr capacity and controled by control valve which was located on the Steam Line. The experimental results show that there is a thermal stratification. Particularly severe thermal plume occured between two spargers rather than other locations, and the temperature on the lower part of water tank was stagnated.

급수가열기 동체 감육 현상 규명을 위한 유동해석 연구

김상녕,김경훈,Kyeong Mo Hwang 한국액체미립화학회 2004 한국액체미립화학회지 Vol.9 No.4

증기발생기 급수 입구 수평배관에서의 열성층 실험 연구

김상녕,김철홍 경희대학교 산학협력기술연구원 2006 산학협력기술연구논문집 Vol.12 No.1

원자력발전소 의 부품 들 은 수명기간동안 열성층 등과 같은 열수력 현상에 의하여 배관 파손 및 이탈,열적 피로,휘어짐 그리고 지지 대 파손등을 경험하고 있다. 이 중 증기발생기 급수 입구부 수평배관에서는 부품의 운전 특성상 빈번하게 열성층이 발생한다. 이에 따라 미국 NRC는 Bulletin 79-13, 88-08 및 88-11 을 통해 주급수배관 및 원자로냉각재계통과 연결된 배관 등 열성층 현상 이 발생 할 것으로 예상되는 배관에 대해 건전성을 입증하도록 요구하고 있다. 한편 증기발생기는 원전의 운전과 안전에서 매우 중요한 부품 중의 하 나 이다 하지만 구형 발전소 에서는 배관 내의 열성 층으로 인한 열응력 이 설계 단계에서 충 분 히 반영되지 못하였다. 따라서 본 연구는 국내 원전 중 충분한 운전 경험이 있고 증 기 발생기 급수 입구부 수평 배관에서 열성층 현상이 일 어 날 가능성이 많은 KSNP 이 전 발전소( 고리 1 ,2,3,4호기 , 영광 1 ,2호기 및 울진 1 ,2호기; 이하에서 이를 구형 발전소 라 칭함)와 KSNP를 대 상으로 보조급수 작동 시 배관 의 크기,보조급수유량,배관 형태 등 여러 운전 조건 에 따른 열성층 및 열주기 발생 가능성 에 대한 실증 실험을 수행하였다. 실험결과는 배 관 내의 열성 층으로 인한 열응력 평가와 방지 대책에 활용될 것이다. Nuclear power plant components suffer pipe shedding, cracking, thermal fatigue, bending and supporting bracket breakage during their life span. Notably, the horizontal inlet nozzle of steam generator is prone to thermal stratification frequently due to its operational characteristics. As a result, PWRs in many countries including the U.S.A. suffered a lot of pipe cracks and leakages around the late 1970s, as the thermal stress inflicted by thermal stratification formed in the horizontal inlet nozzle of steam generator during transition (auxiliary feedwater injection) was not reflected on power plant design. Therefore, we classified the nuclear power plants in Korea into KSNP and Westinghouse-type(W) pawer plants (Kori # 1, 2, 3, 4, Yeonggwang # 1, 2) and Uljin # 1, 2) and conducted an experiment on thermal stratification and thermal cycling in relation to the volume of auxiliary feedwater injected into the horizontal inlet nozzle of steam generator and hot water flowing back from steam generator. As a result, it was found out that KSNP was hardly prone to thermal stratification while thermal stratification occurred in the horizontal inlet nozzle of steam generator in Westinghouse- type(W) power plants, necessitating a solution to address such a phenomenon.

CANDU형 원자로 액체영역제어계통의 불안정 현상 원인 연구

김상녕,고병만,지석준 경희대학교 산학협력기술연구원 2004 산학협력기술연구논문집 Vol.10 No.1

When reactivity insertion such as refueling occurs in CANDU reactors, the power and the water level are tilted in the upper outer zone of LZCS(Liquid Zone Control System) and fluctuate unstably for a certain period of time (1-5 days). The instability described in the above is observed in most of the CANDU reactors in service around the world, but neither its root cause is identified yet nor solutions against it are established. Therefore, this study experimentally and analytically attempted to prove that the root cause lies in the hold-up of light water on the top of TSP(Tube Support Plate) due to the mismatch between the net volumetric flow rate of light water and helium crossing the narrowed porous TSP installed within the LZCS compartment by performing hydrodynamic simulation of in/outflow of light water and helium. And two solutions against the aforementioned instability of LZCS. were suggested. One is to regulate volumes of helium gas flowing into the compartment and light water flowing therefrom and the other is to enlarge flowing paths of helium and light water within TSP. The former may be applicable to nuclear reactors in service and the latter to those planned to be constructed.

피동형 원자로의 Hydraulic Valve 특성 실험

김상녕,김융석,Kim, Sang-Nyung,Kim, Yoong-Seock 대한기계학회 1998 大韓機械學會論文集B Vol.22 No.8

A kind of three-way check valve, so called hydraulic calve was proposed for the substitute of the density lock of passive reactor such as SPWR (System-Integrated Pressurized Water Reactor). The function of the valve are the separation of the borated water from main coolant loop for normal operation and the insurge of the water into the loop for shutdown and the removal of the decay power when the main coolant flow rate is not enough. To verify the operability and the characteristics of the valve, experimental works were executed with 1/3 scale model calve. Also a diffuser model was proposed for the theoretical analysis of the valve.

배관내에서 자유수면 와동에 관한 연구

김상녕,이종원,오율권,장완호 慶熙大學校 1991 論文集 Vol.20 No.-

During mid­loop operation of nuclear power to prevent the residual heat removal system from failure due to air entrainment of free surface vortex in the piping system a set of simulating experiments was performed. Through these experiments, a relation between the non­dimensionalized numbers, such as submergence H/d. Froude number, Reynolds number, was found. However, the effect of the Reynolds number is not so dominant as that of the Froude number. It was also found that the boundary conditions of the free surface, that is, the perturbation of the free surface due to opening of closing pump and valves have significant effects on the free surface vortex. Futhermore a modified inlet device which reducer type with vortex breaker has strongly been recommended for the prevention of air entrainment.

IRWST 수조 내의 온도 분포 실험

김상녕,윤기훈 경희대학교 산학협력기술연구원 2006 산학협력기술연구논문집 Vol.12 No.1

The IRWST (The In-Containment Refueling Water Storage Tank), which is an advanced design concept adopted for APR-1400, has a function to condense the high enthalpy fluid discharged from the Reactor Cooling System (RCS) during a reactor transient. Steam condensation increases the temperature of the coolant in the IRWST, and if the temperature of the water storage tank exceeds 93.3 0C, bubble escape temperature, serious oscillation will happen. This phenomenon would apply mechanical load to the structure of IRWST. Accordingly, an experimental water storage tank was manufactured to simulate the temperature distribution in IRWST when the reactor is transient, and the one was fabricated with a height and volume ratio is 1:1 and 1:400 respectively. As a result of experiments, there was the highest temperature wherein water surface is placed between two spargers, and the temperature of the lower part of water storage tank stagnates at certain point in time. AIso, when the temperature around the sparger reaches the bubble escape temperature, serious oscillation and vibration occurs on the sparger. APR-1400 에 적용되는 설계 개선 사항 중 하나는 격납건물 외부에 설치된 재장전수조 (RWST) 를 격납용기 내부에 설치한 내 격납 재장전수조 (IRWST) 의 채택 이다. IRWST는 원자로 과도 상태시 안전주입모드에서 재순환모드로 전환 시 운전원의 개입 필요성이 없으며, 가압기 안전밸브 혹은 안전 감압밸브 작동 시 방출되는 고온 고압의 유체를 IRWST 내에서 저온수로 응축시켜 격납건물의 방사능 오염을 방지할뿐 아니라 중대시고시 원자로 공동에 냉각수를 공급하는 수원의 역할을 하게 된다. 한편,IRWST 설계 요건은 SDS 작동 시,수조내의 수온이 기포 이탈 온도 (93.3 ℃) 이하를 유지하도록 요구하고 있다. 이러한 온도의 제한은 수조 온도 상승으로 Sparger 에 서 방출된 기포가 응축되지 않고 수면 위로 그대로 상승 하는 것을 방지하기 위함이다 또한 수조 온도 가 포화온도 근방까지 상승하면 수조로 방출되는 증기가 응축진동을 일으키므로 IRWST 에 기계적 하중을 줄 수도 있다. 특히 IRWST 의 Sparger는 비대칭으로 배열되어 있으므로 수직 방향 및 수평방향에 대한 온도 분포가 평탄하지 않을 것으로 예상되며,이는 국부온도를 조기에 기포 이탈 온도로 상승시켜 IRWST 수조의 냉각효율을 저하시킬 수 있다.따라서 본 실험에서는 실제 발전소 내 IRWST의 높이비1:1, 부피 비 1:400 크기 의 환형 수조를 제작하여 증기 응축 시 수조 내 온도 분포와 수조수 온도에 따른 증기 응축 형태를 관찰하였다.

단일 수로내의 상반류 유동 한계

金相寧,梁碩珍,趙相鎭 慶熙大學校 1989 論文集 Vol.18 No.-

A set of experiments of Counter - Current Flow Limit ( CCFL or Flooding ) was executed to improve the drawbacks of Wallis' Correlation which neglects channel length, injection method of liquid and gas phase. In these experiments, using water and air, the followings are found ; ⅰ) The effect of channel length is quite significant, ⅱ) By increasing gas flow rate ( gas velocity ) the flooding point moves upward. This phenomenum is believed to be the effect of flow developing along the channel. ⅲ) The effect of water inlet device is not so significant as that of channel length. But the inlet boundary conditions can also effect the flow developing and flooding afterward, ⅳ) When flooding front reaches the inlet of water injection device, a new reduced flow conditions are set up and resulted in another flooding. This event can be explained by the flooding at the water inlet, therefore, the water can not flow down easily.