• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.636-643
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• 2001

Interference between the crossover leg of the reactor coolant system (RCS) and the pipe whip restraints (PWR) has brought a degradation issue of the integrity of the Reactor Coolant System in Westinghouse type nuclear power plants (NPPs) of Korea. According to the gap Inspect ion carried out during planned overhaul (Year 2000), interference between the crossover leg and the PWR was found in each RCS loop. This plant has had the high vibration problem on the RC pump 'B'. The reason for the high vibration in the RC pump 'B' had been massively surveyed and it was found that the crossover leg of RCS contacted with the PWR in hot condition. Since the contact between the crossover leg and the PWR changes the dynamic characteristics of the piping system for the RCS, this is considered as one reason for the high vibration. And a possibility of overstress on the crossover leg due to the contact with the PWR should be evaluated. Through performing RCS integrity analyses, subsequent actions were initiated to increase the gap between those parts. As the results of the appropriate separation between two parts, it was reported that there was no unusual noise or vibration during plant heat-up. In this paper, the evaluations for the gap between the crossover leg and the PWR and the structural integrity due to loop binding is described.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4685-4694
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• 2023

The ability to calculate the material density sensitivity coefficients of power with respect to the material density has broad application prospects for accelerating Monte Carlo-Thermal Hydraulics iterations. The second-order material density sensitivity coefficients for the general Monte Carlo score have been derived based on the differential operator sampling method in this paper, and the calculation of the sensitivity coefficients of cell power scores with respect to the material density has been realized in continuous-energy Monte Carlo code RMC. Based on the power-density sensitivity coefficients, the sensitivity coefficients of power scores to some other physical quantities, such as power-boron concentration coefficients and power-temperature coefficients considering only the thermal expansion, were subsequently calculated. The effectiveness of the proposed method is demonstrated in the power-density coefficients problems of the pressurized water reactor (PWR) moderator and the heat pipe reactor (HPR) reflectors. The calculations were carried out using RMC and the ENDF/B-VII.1 neutron nuclear data. It is shown that the calculated sensitivity coefficients can be used to predict the power scores accurately over a wide range of boron concentration of the PWR moderator and a wide range of temperature of HPR reflectors.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1079-1087
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• 2018

There are many parameters that affect the natural circulation flow, such as height difference, heating power size, pipe diameter, system pressure and inlet temperature and so on. In general analysis the heating power is often regarded as a uniform distribution. The ANSYS-CFX numerical analysis software was used to analyze the flow heat transfer of supercritical water under different heating power distribution conditions. The distribution types of uniform, power increasing, power decreasing and sine function are investigated. Through the analysis, it can be concluded that different power distribution has a great influence on the flow of natural circulation if the total power of heating is constant. It was found that the peak flow of supercritical water natural circulation is maximal when the distribution of heating power is monotonically decreasing, minimal when it is monotonically increasing, and moderate at uniform or the sine type of heating. The simulation results further reveal the supercritical water under different heat transfer conditions on its flow characteristics. It can provide certain theory reference and system design for passive residual heat removal system about supercritical water.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1999년도 학회창립 10주년 기념 1999년도 가을 학술발표회 논문집
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• pp.329-332
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• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as underground box structure, mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The objective of this paper is largely two folded. Firstly we introduce the cracks control technique by employing low-heat cement mix and thermal stress analysis. Secondly it show the application condition of the cracks control technique like sidewall of LNG in Inchonl.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1999년도 봄 학술발표회 논문집(I)
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• pp.293-298
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• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as underground box structure, mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The objective of this paper is largely two folded. Firstly we introduce the cracks control technique by employing low-heat cement mix and thermal stress analysis. Secondly it show the application condition of the cracks control technique like the subway structure in Seoul.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• 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.