Steam generator tube rupture

Operator failure to depressurize during steam generator tube rupture... 

Woods, D. D. (1982). "Operator Decision Behavior during the Steam Generator Tube Rupture at the Ginna Nuclear Power Station." Research Report 82-1057-CONRM-R2, Westinghouse Research and Development Centre Pittsburgh, PA. 

Cozzuol, J. M., O. M. Hanner, and G. G. Loomis, 1978, Inv. of Influence of Simulated Steam Generator Tube Ruptures during L-O-C Experiments in the Semiscale MOD-1 System, NUREG/CR-0175, EGG Idaho Inc. Idaho Nat. Eng. Lab., Idaho Falls, ID. (4)... 

INPO (1982), Analysis of Steam Generator Tube Rupture Events at Oconee and Ginna, Report 83-030, Institute of Nuclear Power Operations, Atlanta. 

Steam Generator Tube Rupture SCS Heat Exchanger Tube Leak Total Loss of FW Flow... 

In the period fi om October to December 2006, 14 crews of Swedish licensed pressurized water reactor (PWR) operators participated in an experiment performed in H AMMLAB (H Alden Man Machine L AB-oratory Braarud et al. 2007). All crews responded to two versions of a steam generator tube rupture (SGTR) scenario. These simulations constituted the empirical reference data for the International HRA Empirical Study (Lois et al. 2008). While the International HRA Empirical Study focuses on comparing observed performance in HAMMLAB simulator trials with HRA analyses predictions, the present paper concentrates on substantive issues of the observed crew performances namely, the interactions between crews, situations and procedures. 

Immediately following the initiation of the steam generator tube rupture, the operators would likely note a decreasing mass condition in the reactor coolant... 

Additionally, the SCS is used in addition to the SG atmospheric steam release capability and the Emergency Feedwater System to cooldown the RCS following a small break LOCA (see Section 6.3). The SCS is also used subsequent to steam and feedwater line breaks, steam generator tube ruptures, and is used during plant startup prior to RCP restart to maintain flow through the core. 

The original test programme, including three AP-600 major design basis transients, small break loss of coolant accident, steam generator tube rupture and main steam line break, was concluded in October 1994. Two new tests were also performed to verify the facility repeatability and the effect of a more severe cooldown. 

Since the steam generators in an integral reactor are located inside the RPV and can be isolated individually (if there are more than one in the design), there is no need to reduce the primary system pressure in the event of a steam generator tube rupture, unlike the situation in current large loop-type PWRs. 

Steam generator tube ruptures are also accommodated. Immediately following the ruptures significant steam generation occurs however, the lead coolant quickly solidifies in the region of the rupture and steam releases approach those due to the flashing of feedwater only. There is no lead-water reaction, and no public safety concern results. 

Direct bypass of the containment (for example, due to a steam generator tube rupture or to an interfacing systems LOCA which discharges outside the containment) and failure of the containment isolation system should be addressed in the analysis. This would normally be included in the definition of the PDSs. 

N/A LOCA (mtcrfacing eg SGTR (Steam Generator Tube Rupture)... 

LOCA (Primary) Loss of Primary Coolant Accident LOCA (Secondary) Secondary Pipe Rupture (water or steam) LOCA (Interfacing) Steam Generator Tube Rupture ATWS Anticipated Transients Without Scram Transients... 

The design pressure between the steam generator and the isolation valves in the secondary circuit is to equal with that of primary circuit. In the case of a steam generator tube rupture, the isolation valves in the secondary circuit are closed immediately to prevent the release of primary coolant to the environment. 

C-4. J. Li, D. Leaver, and J. Metcalf, Aerosol Retention During An Unisolated Steam Generator Tube Rupture Severe Accident Event , Proceedings on the Fifth International Topical Meeting On Nuclear Thermal Hydraulics, Operations and Safety, Beijing, China, April 13-16, 1997. 

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