• Computational Structural Engineering
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• v.21 no.2
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• pp.41-45
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• 2008

• Computational Structural Engineering
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• v.21 no.2
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• pp.35-40
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• 2008

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.504.1-504.1
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.504.2-504.2
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• 2004

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.9-16
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• 1999

A 330 MWt-sized multi-purpose integral-type reactor, SMART is under development in Korea for the use of nuclear energy other than electricity generation. In this study, various radioactive source terms are estimated for SMART. SMART is different from conventional reactor concepts in operation and design. Therefore Specific Calculation method namely recurrence model is used. This model is based on the change rate in the RC radioactivity materials and operational characteristics of SMART Calculation results show tremendously increase of the levels of RC activity because no cleanup of RC and long term operation.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• Journal of Radiation Protection and Research
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• v.30 no.1
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• pp.39-44
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• 2005

A new developing reactor isn't fixed the structure and the materials of reactor components. To perform the shielding analysis for a reactor vessel by $R-\theta$ geometry, it takes much effort and time to modeling of source term according to the change of reactor components every time. Therefore, we developed the shielding analysis system for the reactor vessel by $R-{\theta}$ geometry, which wasn't affected by the reactor core geometry. By using the developed shielding analysis system, we performed the shielding analysis for the reactor vessel of an integral reactor which has the hexagonal geometry of nuclear fuel assemblies in reactor core. We compared the results obtained from the developed system with those obtained from MCNP analysis. Because the results of developed shielding analysis system were more conservative than those of MCNP calculation, it is useful for shielding analysis. As we had developed the new shielding analysis system for a reactor vessel by $R-{\theta}$ geometry, we reduced error of model for reactor core which was formerly designed by hand and saved the time and the effort to design source term model of reactor core.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4125-4133
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• 2022

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.343-352
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• 2006

Convenient analytical tools for evaluation of the aperiodic and the fluctuating instabilities of the passive residual heat removal system (PRHRS) of an integral reactor are developed and results are discussed from the viewpoint of the system design. First, a static model for the aperiodic instability using the system hydraulic loss relation and the downcomer feedwater heating equations is developed. The calculated hydraulic relation between the pressure drop and the feedwater flow rate shows that several static states can exist with various numbers of water-mode feedwater module pipes. It is shown that the most probable state can exist by basic physical reasoning, that there is no flow rate through the steam-mode feedwater module pipes. Second, a dynamic model for the fluctuating instability due to steam generation retardation in the steam generator and the dynamic interaction of two compressible volumes, that is, the steam volume of the main steam pipe lines and the gas volume of the compensating tank is formulated and the D-decomposition method is applied after linearization of the governing equations. The results show that the PRHRS becomes stabilized with a smaller volume compensating tank, a larger volume steam space and higher hydraulic resistance of the path $a_{ct}$. Increasing the operating steam pressure has a stabilizing effect. The analytical model and the results obtained from this study will be utilized for PRHRS performance improvement.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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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.