The Steam Generator Tube Rupture at Indian Point 

What happened?

On the evening of February 15, the Indian Point 2 reactor was at 99% power. The operators received indication of a steam generator tube rupture. Indications likely were:

Water started flowing from the reactor cooling system (pressurized at 2250 psi) through a small hole in a steam generator tube to the secondary side in the steam generator. On the outside of the tube, water pressure is about 750 psi. Because of the pressure difference and the hole size, radioactive water flowed at a rate of about 90 gallons per minute. As a result, pressurizer level started dropping, exceeding the makeup flow provided by the 98 gpm charging pump. The charging pump provides 3 gpm to the seals of each of the 4 reactor coolant pumps (RCPs). The remainder (~98 - 12) of about 86 gpm would go to the reactor cooling system, if the letdown function of the Chemical & Volume Control System is isolated.

The operators placed 2 charging pumps in service to restore pressurizer level to the normal band; however the pumps could not keep up with the leak, and the operators manually "tripped the unit." This action inserted all control rods shutting down the nuclear reaction in about 2-3 seconds. Nuclear power drops off rapidly at first, then at a -0.3 decades per minute rate, so that within 20 minutes, the reactor power is at 0.000001 % power. Decay heat is still produced after the reactor is shutdown. After the reactor trip, the leak appeared to stop. Later, the leak reinitiated. The operators then followed their emergency operating instructions (E0 - Reactor Trip / Safety Injection and E3 - Steam Generator Tube Rupture). Radiological release was minimized by directing the air ejector vent to a path routed through a particulate-absolute-charcoal (PAC) filter. 

The Con Edison staff declared an Alert, the second lowest emergency action level. The sirens were not sounded. This is only done if there is the potential for a significant release of radioactive materials as could possibly occur in the event of a site area or general emergency. The Con Edison staff notified the NRC and state agencies as required by the plant emergency plan. 

Once the reactor was shutdown, the operators acted to reduce the flow across the steam generator tube. To do this, the operators used a spray valve the pressurizer to reduce the pressure in the reactor cooling system. This could be done because reactor coolant pumps were still available for cooldown (and for spray).  Following the trip, the pressure on the outside of the tube increased to 900 to 1000 psi. These 2 effects reduced the pressure difference across the tube.

The operators then started a cooldown of the reactor cooling system to 200 degrees Fahrenheit by cooling down the three steam generators that did not have the leak. Cooldown is accomplished by using steam dump to condenser from the unaffected steam generators until the reactor coolant system reaches about 300F and 350 psi. At that point, the RHR (Residual Heat Removal) system is started and used for the remainder of the cool down process. The reactor coolant pumps are then shutdown. 

The NRC monitored the event from notification to exit from the alert less than 24 hours later. The plant was shutdown for some time.


Since the early days of commercial nuclear power, plant staffs at all plants have monitored the integrity of the steam generator tubes using eddy current techniques during the normally scheduled refueling outages. Originally, a percentage of the tubes were inspected. If there is a thinning above a preset criterion or indications of abnormalities in the tubes, the tubes would be plugged or sleeved. Plugs prevent water from flowing through the tube. Sleeves may be used in those cases where defects or thinning might be present close to the tube sheet. EPRI has supported industry efforts to improve steam generator reliability. A number of utilities have replaced their original steam generators. It is expected that during the 40 year initial operating license period, a plant may have to replace steam generators once. Excellent chemistry control on the secondary side of the steam generator is required.

This is the first incident of this type since 1993. The estimated 90 gallons per minute for a 4 loop PWR reactor is lower than some earlier events that happened in the 1970's and 1980's. The list below identifies plants that have experienced major steam generator tube leakage events. Those cases indicated by DARK ORANGE are considered as Rupture, where the leak exceeded the charging pump capacity, but was well within the ECCS pump capacity. A Moderate Leak  is less than the charging pump capacity.  Leaks (Graphic shows industry 1990-2000 trend on leaks) in the range of the Technical Specification limits of 1 gpm per steam generator / 500 gallons per day or less. In none of the cases listed or the Indian Point 2 case should there have been a threat to the public health and safety. Data obtained from public data sources, usually the Licensee Event Reports, or LERs).

Year Plant Location Flow Rate Rupture or Leak
1975 Point Beach 1 Wisconsin 125 gal/min Rupture
1976 Surry 2 Virginia 330 gal/min Rupture
1979 Prairie Island 1 Minnesota 390 gal/min Rupture
1982 Ginna New York 630 gal/min Rupture
1987 North Anna 1 Virginia 600 gal/min Rupture
1989 McGuire 1 North Carolina 500 gal/min Rupture
1993 Palo Verde 2  Arizona 240 gal/min Rupture
2000 Indian Point 2  New York 90 gal/min ?? Rupture

The hole that would produce a 90 gallon per minute leak is less than an inch in diameter. More precisely, about 3 inches long by less than 1/4 inch wide.

A steam generator tube rupture is an infrequent event. The NRC Information Notices listed below provide more background on prior events.

Operator initial and requalification training involves training on a plant specific simulator for events of this type. Plant emergency operating procedures address events of this type.

Radioactive Releases 

Radioactive releases can occur early in the event until the steam generator with the broken tube is isolated. Release paths include - air ejector exhaust and turbine driven emergency feed water pump exhaust normally; releases from the steam generator power operated relief valves may occur. The amount released depends on:

The NRC has placed Technical Specification Limiting Conditions for Operations (LCO) limits on: 

These limits are imposed to ensure that, in the event of a steam generator tube rupture, the amount released would have no offsite effects. 

The safety analyses used for licensing the plant assume much higher levels than the NRC allows plants to actually operate with.

The actual release in this event can be expected to be considerably less than a number of prior similar events.

Radioactive release estimates are bounded by the calculations provided in nuclear plant's Updated Safety Analysis Report. These calculations typically assume complete tube failure with a leak rate of over 600 gpm, ongoing leakage prior to the break (e.g. 1 gpm), and 1% defective fuel. 

The radioactive water that leaks into the steam generator must be removed and processed. In tube rupture cases, the water is transferred through the steam generator blowdown system to a Liquid Waste Processing system, which includes storage tanks, ion exchangers, and radiation monitors. Any releases via the waste processing system must meet the requirements of Title 10 of the Code of Federal Regulations, Part 20.

Events prior to the rupture 

Prior leakage in the steam generators was about 3 gallons per day, considerably less than the technical specification limit. 

The New York Times reported that Consolidated Edison has had replacement steam generators on site for some time. These were evidently part of a settlement from a suit against Westinghouse.

Events subsequent to the rupture 

The NRC Augmented Inspection Team completed their visit to the site. NRC held several public meetings related to the event. NRC developed a website related to the event, which provided background information, news and correspondence, and references. The NRC does provide websites containing plant-specific information of interest to the local and general public.

Indian Point's Role in the New York State Energy Production

The Energy Information Administration (EIA) provides detailed information about each state's energy profile. In New York state, Indian Point 2 provides about 2.6 % of the state's total electrical capacity. While, in NY state, nuclear power represents only 16.1% of the capacity, it provides 33.8% of the generated energy. Thus Indian Point 2, in actuality, provides about 5 % of total electrical energy generated in the state. Inspection of the tables in the EIA report shows that NY state has the 2nd highest electricity costs in the nation. If gas and oil were used at levels higher than they currently are, consumers' electricity costs would be even higher than they currently are. When oil and gas costs rise, one could expect electricity costs to follow the same trend.

Ownership Change

Since the 2000 event, Consolidated Edison and New York State Power Authority sold Indian Point Units 2 and 3 to Entergy. The Entergy-Nuclear division is responsible for operation of the reactors to supply the wholesale market.


Copyright 1996-2006. The Virtual Nuclear Tourist. All rights reserved. Revised: December 24, 2005.