• 한국압력기기공학회 논문집
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• 제5권1호
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.42-47
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• 2012

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.

• 한국압력기기공학회 논문집
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• 제8권3호
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• pp.1-6
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• 2012

The containment Integrated Leakage Rate Testing(ILRT) of nuclear power plants in Korea is performed in accordance with NSSC(Nuclear Safety and Security Commission) code 2012-16 and ANSI/ANS 56.8-1994. Nuclear power plants in Korea and the United States are to apply same test criteria, ANSI/ANS 56.8-1994, except type A testing time. NPPs in Korea apply 24 hours according to NSSC code 2012-16, but NPPs in United States apply 8 hours according to 10CFR50 App. J for type A test. So, there are many difficulties in order to perform ILRT in Korea. In this study, I review the impact on the ILRT results and the effect of ILRT due to type A testing time. The future, we will continue study to enhance the test reliability and improve these problems.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.617-648
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• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• 한국안전학회지
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• 제35권3호
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• pp.96-104
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• 2020

The numerical simulations were conducted to investigate the thermal-fluid phenomena occurred inside the experimental apparatus during a PCCS, used to remove heat released in accidents from a containment of light water nuclear power plant, operation. Numerical simulations of the flow and heat transfer caused by wall condensation inside the containment simulation vessel (CSV), which equipped with 18 vertical heat exchanger tubes, were conducted using the commercial computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the wall condensation model were used for turbulence closure and wall condensation, respectively. The simulation using the actual size of the apparatus. However, rather than simulating the whole experimental apparatus in consideration of the experimental cases, calculation resources, and calculation time, the simulation model was prepared only in CSV. Selective simulation was conducted to verify the effects of non-condensable gas(NC gas) concentration, CSV internal pressure, and wall sub-cooling conditions. First, as a result of the internal flow of CSV, it was observed that downward flow due to condensation occurred surface of the vertical tube and upward flow occurred in the distant place. Natural convection occurred actively around the heat exchanger tube. Due to this rising and falling internal flow, natural circulation occurred actively around the heat exchanger tubes. Next, in order to check the performance of built-in condensation model using according to the non-condensable gas concentration, CSV internal flow and wall sub-cooling, the heat flux values were compared with the experimental results. On average, the results were underestimated with and error of about 25%. In addition, the influence of CSV internal pressure and wall sub-cooling was small, but when the condensate was highly generated due to the low non-condensable gas concentration, the error was large compared to the experimental values. This is considered to be due to the nature of the condensation model of the CFX code. However, in spite of the limitations of CFD, it is valid to use the built-in condensation model of CFD for PCCS performance prediction from a conservative perspective.

• 전산구조공학
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• 제8권1호
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• pp.4-12
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• 1995

원자로 격납건물(Reactor Containment Bldg)은 정상가동시는 물론 냉각재상실사고(LOCA)를 포함하는 설계기준사고(DBA) 및 설계기준지진(DBE) 발생시 구조물 자체의 건전성 확보는 물론 주기기(NSSS Equipment)를 포함하는 안전관련 계통 및 기기를 안전하게 보호/지지하므로써 핵누출을 방지하여 발전소 종사자를 포함하는 국민의 재산과 생명을 보호하는 역할을 하는 원자력발전소에서 가장 중요한 구조물이다. 원자로 격납건물은 압력용기(Pressure Vessel : 설계내압 5 psi 이상인 용기)로 설계되는 격납용기와 1, 2차 차폐구조 등의 내부구조물로 구성되는데 이 중 본 소고에서는 격납용기의 해석 및 설계 그리고 구조건전성 시험 및 사용중검사에 대해서만 간략하게 기술한다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.

• 태양에너지
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• 제4권2호
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• pp.43-49
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• 1984

Experiments were performed to study freezing on a finned vertical tube when either conduction in the solid or natural convection in a liquid controls the heat transfer. Conduction is the controlling mode when the liquid is at its fusion temperature, whereas natural convection controls when the liquid temperature is above the fusion value. The liquid was housed in a cylinderical containment vessel whose surface was maintained at a uniform, time-invariment temperature during a data run, and the freezing occurred on a finned vertical tube positioned along the axis of the vessel. The phase change medium was n-octacosan, a paraffin which freezes at about $61^{\circ}C$. For conduction-controlled freezing, the enhancement of the frozen mass due to finning is greatest when the frozen layer is thin and decrease as the layer grows thicker. The degree of enhancement is generally less than the surface area ratio of the finned and unfinned tube.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3485-3490
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• 2007

The TEXAS-V code tuned for TROI-13 was used for analyzing the parametric findings in TROI experiments. The calculations on the melt composition are relatively similar to the TROI experimental results. The water depth effect in TEXAS-V code seems to be consistent with TROI experiments in some degree. The water area effect of TEXAS-V calculations seems not to be harmonious to that in TROI experiments. This seems to indicate that TEXAS-V as 1-dimensional code or as the numerical steam explosion has a limitation on estimating area effect. Thus, TEXAS-V tuned for TROI-13 seems to have an ability to estimate the parametric effect of TROI experiments. The evaluated TEXAS-V was used for estimating the ex-vessel steam explosion load. The calculated explosion pressure and load were about 40 MPa and 75 kPa.sec, which are not much threatening level for containment integrity, but are arguable value for the integrity.