• Nuclear Engineering and Technology
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• 제39권4호
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• 대한기계학회논문집
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• 제19권6호
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• pp.1451-1459
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• 1995

The objective of this paper is to evaluate the material properties of SA312 TP316 and SA312 TP304 stainless steels and their associated welds manufactured for safety injection system of Yonggwang 3,4 nuclear generating stations. A total of 62 tensile tests and 46 fracture toughness tests were conducted and the effects of various parameters such as pipe size, crack plane orientation, tests were conducted and the effects of various parameters such as pipe size, crack plane orientation, test temperature, welding on material properties were discussed. Test results show that the effect of test temperature on fracture toughness was significant while the effects of pipe size and crack plane orientation on fracture toughness were negligible. Fracture toughness of the weld metal was in general higher than that of the base metal.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.283-288
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• 1996

Direct safety injection into the reactor vessel downcomer annulus(DVI) is a fundamental feature of the KNGR(Korean Next Generation Reactor) four-train safety injection system. The numerical analysis of thermal mixing of ECC(Emergency Core Cooling) water through DVI with the water in the RVDC(Reactor Vessel Downcomer) annulus has been performed, in order to study the impact of nozzle location on the pressurized thermal shock and safety analysis. The results of this study show that the thermal mixing due to the natural circulation induced by the limiting accident conditions is sufficient to prevent temperature in the RVDC from dropping to the level of concern for PTS. When the DVI nozzle is located right above the cold leg, the temperature distribution at the outlet of flow field is most uniform. The tool used for numerical analysis is CFDS-FLOW3D.

• 한국정밀공학회지
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• 제28권12호
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• pp.1372-1378
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• 2011

The smart air condition system is superior to conventional air condition system in the aspect of control accuracy, environmental preservation and it is foundation for intelligent vehicle such as electric vehicle, fuel cell vehicle. In this paper, failure analyses of smart air condition system will be performed and then sensor fusion technique will be proposed for fail safety of smart air condition system. A sensor fusion logic of air condition system by using CO sensor, $CO_2$ sensor and VOC, $NO_x$ sensor will be developed and simulated by fault injection simulation. The fusion technology of smart air condition system is generated in an experiment and a performance analysis is conducted with fusion algorithms. The proposed algorithm adds the error characteristic of each sensor as a conditional probability value, and ensures greater accuracy by performing the track fusion with the sensors with the most reliable performance.

• 대한기계학회논문집B
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• 제41권10호
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• pp.675-683
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• 2017

본 연구에서 원자력 발전용 비상노심냉각 장치인 안전주입탱크의 재충수 단계에 대한 유량에 대한 이론해석을 수행했다. 이론해석을 통해 재충수 단계 유동에 대한 이론 모형을 정립하고 재충수 단계에 대한 비선형 유량방정식을 구하고, 테일러 급수 전개법을 통해 근사유량방정식과 냉각수의 자유표면 높이변화와 유량변화를 예측할 수 있는 이론해들을 구했다. 기존연구에 나와 있는 실험과 비교하여 이론해의 유용성을 검증했다.

• Journal of Mechanical Science and Technology
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• 제17권9호
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• 한국압력기기공학회 논문집
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• 제7권1호
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• pp.1-8
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• 2011

Advanced SITs of the evolutionary PWRs have the advantage that they can passively control the ECC water discharge flow rate. Thus, the LPSI pumps can be eliminated from the safety injection system owing to the benefit of the advanced SITs. In the present study, a passive sealing plate was designed in order to overcome the shortcoming of the advanced SITs, i.e., the early nitrogen discharge through the stand pipe. The operating principle of the sealing plate depends only on the natural phenomena of buoyancy and gravity. The performance of the sealing plate was evaluated using the VAPER test facility, equipped with a full-scale SIT. It was verified that the passive sealing plate effectively prevented the air discharge during the entire duration of the ECC water discharge. Also, the major performance parameters of the advanced SIT were not changed with the installation of the sealing plate.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.979-988
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• 2017

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.180.1-180
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.176-177
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• 1999