• 대한기계학회논문집A
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• 제30권10호
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• pp.1173-1178
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• 2006

In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and $330^{\circ}C$, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at $430^{\circ}C$, respectively. The specimens for elastic-plastic fracture toughness tests are according to the process in the thermal notch is created in the heat affected zone(HAZ) of CF8M and deposited zone. From the experiments, the $J_{IC}$ value notched in HAZ of CF8M presented a rapid decrease up to 300 hours at $430^{\circ}C$ and slowly decreased according to the process in the thermal aging time. Also, the $J_{IC}$ value presented a lower value than that of the CF8M base metal. And, the $J_{IC}$ of the deposited zone presented the lowest value of all other cases.

• 대한기계학회논문집B
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• 제34권12호
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• pp.1093-1099
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• 2010

초고온원자로(VHTR) 설계에 있어 중간열수송루프(IHTL) 및 중간열교환기(IHX) 설계는 고온의 운전조건($950^{\circ}C$ 이상)으로 인하여 공학적으로 어려운 과제 중 하나로 알려져있다. 본 연구에서는 LiF, NaF 및 KF(46.5:11.5:42.0 mole %)의 공융혼합물인 Flinak molten salt 를 IHTL 의 열수송매체로 고려하였다. Flinak molten salt 의 세관에서의 열수력 특성을 평가하기 위하여 직경이 1.4 mm 인 원형관을 이용하여 고온의 가스와 Flinak 을 열교환할 수 있는 이중관식 열교환기를 구성하여 실험하였다. 실험 결과 층류유동에서 측정된 Flinak 의 마찰계수는 이론식인 64/Re 에 근접하였고 Nusselt 수는 일반적으로 3.66 에서 4.36 범위에 들었다.

• 대한기계학회논문집B
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• 제32권11호
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• pp.897-905
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• 2008

Selective catalytic reduction(SCR) is known as one of promising methods for reducing $NO_x$ emissions in diesel exhaust gases. $NO_x$ emissions react with ammonia in the catalyst surface of SCR system at working temperature of catalyst. In this study, to raise the reacting temperature when the exhaust gas temperature is too low, a heater is located at the bottom of SCR reactor. At an ambient temperature, ammonia is radially injected perpendicular to the exhaust gas flow at inlet pipe and uniformly mixed in the mixing area after being impinged against the wall. To predict the turbulent model inside the mixing area of SCR system, the standard ${\kappa}\;-\;{\varepsilon}$ model is applied. This work investigates numerically the effects of induced heat on the gaseous flow. The results show that the Taylor-$G{\ddot{o}}rtler$ type vortex is generated after the gaseous flow impinges the wall in which these vortices influence the temperature distribution. The addition of heat disturbs the flow structure in bottom area and then stretching flow occurs. Vorticity strand is also formed when heat is continuously increased. Constriction process takes place, however, when a further heat input over a critical temperature is increased and finally forms shed vortex which is disconnected from the vorticity strand. The strong vortex restricts the heat transport in the gaseous flow.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1436-1445
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• 2021

To obtain the dynamic characteristics of reactor secondary circuit under transient conditions, the system analysis program was developed in this study, where dynamic models of secondary circuit were established. The heat transfer process and the mechanical energy transfer process are modularized. Models of main equipment were built, including main turbine, condenser, steam pipe and feedwater system. The established models were verified by design value. The simulation of the secondary circuit system was conducted based on the verified models. The system response and characteristics were investigated based on the parameter transients under emergency shutdown and overload. Various operating conditions like turbine emergency shutdown and overspeed, condenser high water level, ejector failures were studied. The secondary circuit system ensures sufficient design margin to withstand the pressure and flow fluctuations. The adjustment of exhaust valve group could maintain the system pressure within a safe range, at the expense of steam quality. The condenser could rapidly take out most heat to avoid overpressure.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3635-3642
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• 2021

Natural circulation systems (NCSs) are extensively applied in nuclear power plants because of their simplicity and inherent safety features. For some passive natural circulation systems in floating nuclear power plants (FNPPs), the ocean is commonly used as the heat sink. Condensation induced water hammer (CIWH) events may appear as the steam directly contacts the subcooled seawater, which seriously threatens the safe operation and integrity of the NCSs. Nevertheless, the research on the formation mechanisms of CIWH is insufficient, especially in NCSs. In this paper, the characteristics of flow rate and fluid temperature are emphatically analyzed. Then the formation types of CIWH are identified by visualization method. The experimental results reveal that due to the different size and formation periods of steam slugs, the flow rate presents continuous and irregular oscillation. The fluid in the horizontal hot pipe section near the water tank is always subcooled due to the reverse flow phenomenon. Moreover, the transition from stratified flow to slug flow can cause CIWH and enhance flow instability. Three types of formation mechanisms of CIWH, including the Kelvin-Helmholtz instability, the interaction of solitary wave and interface wave, and the pressure wave induced by CIWH, are obtained by identifying 67 CIWH events.

• International Journal of Air-Conditioning and Refrigeration
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• 제11권2호
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• pp.51-60
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• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. Standard k-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• 한국안전학회지
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• 제30권3호
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• pp.107-113
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• 2015

The goal of this paper is to investigate structural integrity factors of RMI(reflective metal insulation) to confirm the design requirements in nuclear power plant. Currently, a glass wool insulation is using now, but it will gradually be replaced with the reflective metal insulation maded by stainless steel plates. The main function of an insulation is to minimize a heat loss of vessel and pipes in RCS(reactor coolant system). It has to maintain structural a integrity in nuclear power plant life duration. In this study, the structural integrity analysis was carried out both multi-plate and outer shell plate by using a static analysis and experimental test. First, inner multi-plate has a self support structure for being air space. Because the effect of total static weight in multi-layer plate is low, a plate collapse possibility is not high. Considering optimum thin plate pressing process, it has to pre-check the basic physical properties. Second, the outer segment thickness and stiffener shape are verified by the numerical static analysis, and sample test for both type of panel and cylindrical pipe model.

• 한국전산유체공학회지
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• 제20권1호
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• pp.39-46
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• 2015

In this research, we studied the pressure drop affecting on the internal surface roughness and the curvature radius of a U-tube, which is used for the cooling system in PWR(Pressurized Water Reactor). Using ANSYS-FLUENT, a commercial code based on CFD(Computational Fluid Dynamics) technique, we compared a Moody chart with the Darcy friction factor changed by a range of various surface roughness and Reynolds numbers of a straight pipe model. We studied the effect giving variation about a range of various surface roughness and the curvature radius of the full scale U-tube model. The material of the heat transfer tube is Inconel 690 used in the steam generator. We compared the velocity distribution of selected 4 locations, and derived the correlation between the surface roughness and the pressure drop for the U-tube of each representative curvature radius using the linear regression method.

• 설비공학논문집
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• 제15권3호
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• pp.239-245
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• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. $textsc{k}$-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of the main flow in the duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from the main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.311-318
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• 1992

원자력 발전소에서 Mid-loop 운전시 배관내에서 발생하는 자유수면 와동으로 인해 잔열 제거계통 배관내 공기가 흡입될 가능성이 있으며 이로 인한 계통상실 방지를 위하여 수위와 흡입유량과의 관계를 실험을 통해서 H/d, 프라우드수, 레이놀즈 수 등과 같은 무차원 수로 구하였다. 실험결과 레이놀즈수는 크게 영향을 미치지 않았으며 주로 프라우드수가 자유수면 와동을 지배하는 것으로 판명되었다. 한편 운전시 펌프나 밸브의 개폐로 인한 수면의 섭동이 와동에 많은 영향을 미치는 것이 밝혀졌다. 원자력 발전소의 안전과 관련하여 배관내에서 와동으로 인한 공기흡입 방지책으로 Reducer형의 흡입구 개선방안을 제시하였다.