Safety depressurization system for Korean next generation reactor

Conference · Mon Dec 01 04:00:00 EST 1997

OSTI ID:552168

Kwon, Young Min; Lim, Hong Sik; Song, Jin Ho ^[1]

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1. Korea Atomic Energy Research Institute (KAERI), Taejon (Korea, Republic of); and others

Korean Next Generation Reactor (KNGR) adopted an advanced design feature of safety depressurization system (SDS) to rapidly depressurize the primary system in case of beyond design basis events. Two design approaches are considered for the SDS design. The use of bleed valves similar to those of ABB-CE`s System80+ is design option 1, while in the design option 2 the French Sebim valve is considered to provide combined function of overpressure protection and rapid depressurization. In this paper thermal hydraulic analyses using a best-estimate version of CEFLASH-4AS/REM are performed for total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option various feed and bleed (F&B) procedures are investigated for TLOFW event. For the design option 1 the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The minimum analytical bleed area to prevent core uncovery is determined as 0.028 ft{sup 2} per bleed path. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with sufficient margin by using design option 1. For the design option 2 the operator action times for initiating the F&B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If operator opens two out of three Sebim valves in conjunction with four HPSI pumps before hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of present investigation demonstrate that the two design options are both feasible. 10 refs., 12 figs., 2 tabs.

• Research Organization: American Nuclear Society, La Grange Park, IL (United States)
• OSTI ID: 552168
• Report Number(s): CONF-970607--Vol.2
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
DEPRESSURIZATION SYSTEMS
DESIGN
LOSS OF FLOW
PWR TYPE REACTORS
SAFETY ANALYSIS
VALVES