Containment event analysis for postulated severe accidents: Sequoyah Power Station, Unit 1

Technical Report · Sun Feb 01 04:00:00 EST 1987

OSTI ID:6614111

Behr, V L; Benjamin, A S; Kunsman, D M; Lewis, S R; Murfin, W B

A study has been performed as part of the Severe Accident Risk Reduction Program (SARRP) to investigate the response of a particular pressurized water reactor with an ice-condenser containment (Sequoyah Unit 1) to postulated severe accidents. A detailed containment event tree for the Sequoyah plant has been devised to describe the various possible accident pathways that can lead to radioactive releases from containment. Data and analyses from a large number of NRC and industry-sponsored programs have been reviewed and used as a basis for quantifying the event tree, i.e., determining the likelihood of each pathway for a variety of accident sequence initiators. A generalized containment event tree code, called EVNTRE, has been developed to facilitate the quantification. The uncertainty in the results has been examined by performing the quantification three times, using a different set of input each time to represent the variation of opinion in the reactor safety community. In the so-called ''central'' estimate, the likelihood of early containment failure (occurring before or at the time of reactor vessel breach) was found to be high for station blackout sequences but very low for other accident sequence initiators. Unavailability of igniters and air return fans was the principal reason for the high failure probability for station blackouts. The analysis also showed that melting or bypass of the ice before or within a short time after vessel breach can be expected to occur with moderate to high likelihood during station blackouts and during sequences initiated by very small LOCAs with failure of emergency core cooling in the recirculation phase after success in the injection phase. This work supports NRC's assessment of severe accident risks to be published in NUREG-1150.

• Research Organization: Sandia National Labs., Albuquerque, NM (USA); Nuclear Regulatory Commission, Washington, DC (USA). Div. of Reactor Systems Safety
• DOE Contract Number: AC04-76DP00789
• OSTI ID: 6614111
• Report Number(s): NUREG/CR-4700-Vol.2; SAND-86-1135-Vol.2; ON: TI87008439
• Country of Publication: United States
• Language: English

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