Concern on accident management for the Korea next generation reactor

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

OSTI ID:552109

Seok, Ho; Lim, Hakkyu; Kang, Sun-Koo ^[1]

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1. Korea Power Engineering Co., Inc., Kyunggi-do (Korea, Republic of); and others

The Korean Next Generation Reactor (KNGR) is under development to be built after year 2000 in Korea. To enhance its capability of preventing and/or mitigating severe accidents, various safety features are incorporated in its design. Some of them are designed against severe accidents and can be operated based on accident management program (AMP) for the KNGR. In this study, the potential capability of the Safety Depressurization System (SDS) and the Shutdown Cooling System (SCS) to mitigate the consequence of severe accidents was examined by using the MAAP 4.02 code as a preliminary step of the AMP development for the KNGR. The concerned accident sequences are small break loss of coolant accidents (SB LOCAs) with a failure of high pressure safety injection system (HPSIS) and a total loss of feedwater (TLOFW). In the level 1 Probabilistic Safety Assessment (PSA) of the KNGR, the operation of the SDS and SCS was not considered because the failures of the HPSIS and the aggressive secondary side cooling result in core damage based on the success criteria of the level 1 PSA. The analysis results show that the SDS can depressurize the RCS below the shutoff head of the shutdown cooling system (SCS) prior to reactor vessel failure. Although core uncovery and core damage occur early due to the opening of the SDS valves, the MAAP calculation results show that the SCS can reflood the damaged core and that core damage and reactor vessel failure can be mitigated or prevented by the feed-and-bleed operation with those systems. From the analysis results, therefore, it seems that the operation of the SDS and SCS can provide a means of mitigating accident consequences and can be employed as an effective accident management strategy for the KNGR. 5 refs., 6 figs., 4 tabs.

• Research Organization: American Nuclear Society, La Grange Park, IL (United States)
• OSTI ID: 552109
• 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
LOSS OF COOLANT
LOSS OF FLOW
M CODES
PWR TYPE REACTORS
REACTOR COOLING SYSTEMS
REACTOR SHUTDOWN
SAFETY ANALYSIS