Candu 6 severe core damage accident consequence analysis for steam generator tube rupture scenario using MAAP4-CANDU V4.0.5A: preliminary results

Conference · Sat Jul 01 04:00:00 EDT 2006

OSTI ID:21021211

Petoukhov, S M; Awadh, B; Mathew, P M ^[1]

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1. Chalk River Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

This paper describes the preliminary results of the consequence analysis for a generic AECL CANDU 6 station, when it undergoes a postulated, low probability Steam Generator multiple Tube Rupture (SGTR) severe accident with assumed unavailability of several critical plant safety systems. The Modular Accident Analysis Program for CANDU (MAAP4-CANDU) code was used for this analysis. The SGTR accident is assumed to begin with the guillotine rupture of 10 steam generator tubes in one steam generator in Primary Heat Transport System (PHTS) loop 1. For the reference case, the following systems were assumed unavailable: moderator and shield cooling, emergency core cooling, crash cool-down, and main and auxiliary feed water. Two additional cases were analyzed, one with the crash cool-down system available, and another with the crash cool-down and the auxiliary feed water systems available. The three scenarios considered in this study show that most of the initial fission product inventory would be retained within the containment by various fission product retention mechanisms. For the case where the crash cool-down system was credited but the auxiliary feed water systems were not credited, the total mass of volatile fission products released to the environment including stable and radioactive isotopes was about four times more than in the reference case, because fission products could be released directly from the PHTS to the environment through the Main Steam Safety Valves (MSSVs), bypassing the containment. For the case where the crash cool-down and auxiliary feed water systems were credited, the volatile fission product release to the environment was insignificant, because the fission product release was substantially mitigated by scrubbing in the water pool in the secondary side of the steam generator (SG). (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21021211
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ACCIDENTS
ATOMIC ENERGY OF CANADA LTD
CANDU TYPE REACTORS
CONTAINMENT
COOLING
DAMAGE
ECCS
FISSION PRODUCT RELEASE
FISSION PRODUCTS
HEAT TRANSFER
INVENTORIES
M CODES
MODERATORS
RELIEF VALVES
RUPTURES
SAFETY
STEAM GENERATORS
TUBES
WATER