Analysis of induced steam generator tube rupture using MAAP 4.0

Journal Article · Tue Dec 31 04:00:00 EST 1996 · Transactions of the American Nuclear Society

OSTI ID:437029

Kenton, M ^[1]; Epstein, M; Henry, R E; Paik, C ^[2]; Fuller, E ^[3]

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1. Dames & Moore, Westmont, IL (United States)
2. Fauske & Assoc., Inc., Burr Ridge, IL (United States)
3. Polestar Applied Technology, Los Altos, CA (United States)

The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes to leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes.

• OSTI ID: 437029
• Report Number(s): CONF-9606116--
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 74; ISSN 0003-018X; ISSN TANSAO
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
99 GENERAL AND MISCELLANEOUS
FEEDWATER
LEAKS
M CODES
PWR TYPE REACTORS
REACTOR ACCIDENTS
RUPTURES
STEAM GENERATORS
TUBES