• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.257.1-257
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• 2001

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.165-175
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• 1995

The overpressure protection system for PWR should be designed with sufficient capacity to limit the pressure to less than 110％ of the reactor coolant system design pressure during the most severe abnormal operational transient. In this study, the feasibility of adopting the SEBIM POSRV instead of the current spring loaded pop-opening safety valves to the ABB-CE designed 2825 MWt PWR is investigated for its overpressure protection capability. The required SEBIM POSRV size as well as its opening/closing setpoints are determined through a series of computer analyses using the LTC code which has been used for the overpressure protection analysis for Yonggwang units 3&4. The analysis results show that the overpressure protection system with monobloc SEBIM POS-RV can maintain the RCS pressure below 110％ of the design pressure demonstrating its overpressure protection capability for the ABB-CE designed 2825 MWt PWRs.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.184-184
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.255-255
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.182.1-182
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• 2001

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.225-236
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• 1994

The sudden discharge of the loop seal water, which is present upstream of the SEBIM POSRV, creates large momentum and inertia forces on the downstream of the discharge piping system. This study provides the procedures and results of analysis of the thermal-hydraulic transient in the SEBIM POSRV discharge piping during the valve opening. The analysis is peformed by RELAP5/MOD3. The appropriate modeling of the discharge piping system, SEBIM POSRV opening characteristics, and loop seal water discharge for the RELAP5/MOD3 analysis is suggested. Also performed is the sensitivity study for the selection of proper options for the junction and volume control. flags. The analysis results demonstrate the adequacy of the RELAP5/HOD3 for the thermal-hydraulic transient analysis of the loop seal water discharge of the SEBIM POSRV discharge piping system. From the sensitivity analysis results, it is shown that the smooth area change option with reasonable geometric pressure drop distribution, non-equilibrium option, and proper time step should be selected for loop seal water discharge analysis.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.568-573
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• 2003

When a safety valve equipped in a nuclear power plant opens in an instant by an accident, a moving shock wave propagates downstream the valve, inducing a complicated unsteady flow field. The moving shock wave may exert severe load to the structure. So, to reduce the load acting on the wall of POSRV, a gradual opening of POSRV is adopted in general. In theses connections, a numerical work is performed to investigate the effect of valve opening time on the unsteady flow fields downstream of the valve. Compressible, two-dimensional Navier-Stokes equations are used with the finite volume method. The obtained results show that sharp pressure rise through moving shock tor the case of instant opening is attenuated by employing the gradual opening of valve. It is turned that the flows for the two cases of gradual valve opening time show the similar to that of highly under-expanded one in jet structure having expansion and compression waves and Mach stem. Also, comparing with the results for the two cases of opening time, the shorter the valve opening is, the pressure gradient at the downstream of the valve becomes softly.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2012.05a
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• pp.717-718
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• 2012

• Journal of Energy Engineering
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• v.12 no.2
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• pp.74-83
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• 2003

In the advanced nuclear power plants including APR1400, the SDVS (Safety Depressurization and Vent System) is adopted to increase the plant safety using the concept of feed-and-bleed operation. In the case of the TLOFW (Total Loss of Feedwater), the POSRV (Power Operated Safety Relief Value) located at the top of the pressurizer is expected to open due to the pressurization of the reactor coolant system and discharges steam and/or water mixture into the water pool, where the mixture is condensed. During the condensation of the mixture, thermal-hydraulic loads such as pressure and temperature variations are induced to the pool structure. For the pool structure design, such thermal-hydraulic aspects should be considered. Understanding the phenomena of the submerged steam jet condensation in a water pool is helpful for system designers to design proper pool structure, sparger, and supports etc. This paper reviews and evaluates the steam jet condensation in a water pool on the physical phenomena of the steam condensation including condensation regime map, heat transfer coefficient, steam plume, steam jet condensation load, and steam jet induced flow.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.27 no.10
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• pp.1464-1471
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• 2003

In order to investigate the flows with shock wave in branch, 108$^{\circ}$ elbow and T-junction of the IRWST system of standard Korean nuclear reactor, detail time dependent behaviors of unsteady flow with shock wave, vortex and so on are obtained by numerical method using compressible three-dimensional Navier-Stokes equations. At first, the complex flow including the incident and reflected shock waves, vortex and expansion waves which are generated at the corner of T-junction is calculated by the commercial code of FLUENT6 and is compared with the experimental result to obtain the validation of numerical method. Then the flow fields in above mentioned units are analyzed by numerical method of [mite volume method. In numerical analysis, the distributions of flow properties with the moving of shock wave and the forces acting on the wall of each unit which can be used to calculate the size of supporting structure in future are calculated specially. It is found that the initial shock wave of normal type is re-established its type from an oblique one having the same strength of the initial shock wave at the 4 times hydraulic diameters of downstream from the branch point of each unit. Finally, it is turned out that the maximum force acting on the pipe wall becomes in order of the T-junction, 108$^{\circ}$ elbow and branch in magnitude, respectively.