Information Notice 2000-09: Steam Generator Tube Failure at Indian Point Unit 2
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
June 28, 2000
| NRC INFORMATION NOTICE 2000-09: | STEAM GENERATOR TUBE FAILURE AT INDIAN POINT UNIT 2 |
Addressees
All holders of operating licenses for nuclear power reactors except those who have ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform addressees of a steam generator tube failure at Indian Point Unit 2. NRC investigations of the licensee's steam generator inspection program are ongoing and any potentially generic issues identified will be communicated in a separate generic communication. However, the investigations to date re-emphasize the importance of licensee involvement with ongoing industry efforts to understand and detect steam generator degradation. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances
On February 15, 2000, at 7:17 p.m., the Indian Point Unit 2 nuclear plant experienced a steam generator tube failure, which required the declaration of an Alert at 7:29 p.m., and a manual reactor trip at 7:30 p.m. The operators identified that the #24 steam generator was the source of the leak and completed isolation of the #24 steam generator by 8:31 p.m.
At 9:02 p.m., the operator opened the high-pressure steam dump valves and established an excessive primary plant cooldown rate that caused a rapid reduction in the pressurizer level and required the operators to manually initiate safety injection. The operators reset the safety injection at 9:21 p.m., reduced the reactor coolant system pressure to about 970 psig at 9:32 p.m., and re-commenced a plant cooldown at 11:35 p.m.
The residual heat removal (RHR) system was placed in service on February 16, 2000, at 12:38 p.m., and primary plant pressure was reduced below the #24 steam generator pressure to terminate the steam generator tube leakage at 2:20 p.m. The plant cooldown continued, and the plant entered cold shutdown at 4:57 p.m. The licensee exited the Alert at 6:50 p.m.
The NRC sent an Augmented Inspection Team (AIT) on February 18, 2000, to review the causes, safety implications, and licensee actions associated with the event. The AIT developed a sequence of events, determined the risk significance of the event, and assessed the response by the plant staff and management. The cause of the tube failure was outside the scope of this inspection and is currently being reviewed separately by the NRC. The AIT's report is presented in Inspection Report 05000247/2000-02, dated April 28, 2000 (Accession Number ML003710036).
Discussion
The event was risk significant. It involved a steam generator tube failure that resulted in an initial primary-to-secondary leak of reactor coolant of approximately 146 gallons per minute and required an "Alert" declaration (the second level of emergency action in the NRC-required emergency response plan). The event resulted in a minor radiological release to the environment that was well within regulatory limits. No radioactivity was measured offsite above normal background levels, and the event did not adversely impact the public health and safety.
The licensee performed the necessary actions to protect the health and safety of the public. Specifically, the operators promptly and appropriately took those actions in the emergency operating procedures to trip the reactor, isolate the affected steam generator, and depressurize the reactor coolant system. Additionally, the necessary event mitigation systems worked properly. Notwithstanding the above actions, the AIT identified performance problems in several broad areas that challenged operators, complicated the event response, delayed achieving the cold shutdown condition, and affected the radiological release. The problems involved operator performance, procedure quality, equipment performance, technical support, and emergency response.
Operator Performance
Some operator performance problems were noted during the plant cooldown phase involving the following:
- While attempting to cool down the reactor coolant system (RCS), the reactor operator initiated an excessive cooldown rate that exceeded procedural and Technical Specification limits. The excessive cooldown led to several conditions that complicated the subsequent event response and delayed the RCS cooldown.
- Operators were slow to recognize configuration lineup problems that (1) prevented successful operation of the auxiliary spray system to lower RCS pressure and (2) delayed heatup of the RHR system.
Procedure Quality
The procedures adequately guided the initial operator response; however, several procedure problems were identified that delayed the cooldown and depressurizing of the RCS. Procedure deficiencies affected Standard Operating Procedures, Emergency Operating Procedures, and Emergency Plan Implementing Procedures. Specific activities included initiation of RHR cooling, initiation of component cooling water alignment, use of auxiliary pressurizer spray, use of methods to monitor RCS temperature to maintain cold shutdown conditions, and initiation of emergency response organization (ERO) notifications. Station personnel were previously aware of the procedure issue involving initiation of RHR cooling but had not corrected the problem before this event.
Equipment Performance
The necessary event mitigation systems, including the reactor protection system, the auxiliary feedwater system, and the safety injection system, functioned properly. However, several longstanding equipment performance problems were identified that challenged operators during this event:
- Two losses of condenser vacuum resulted from problems with the operation of the automatic steam supply pressure control valve to the steam jet air ejectors, and the #22 condenser vacuum pump.
- The isolation valve seal water system became inoperable during the event and required operator action and an entry into a Technical Specification Limiting Condition for Operation Action Statement.
- A containment entry was required to install a temporary nitrogen supply to the pressurizer power-operated relief valve to compensate for a design deficiency. This action was required before placing the overpressure protection system in service.
- The steam generator leak rate monitoring equipment had been degraded for an extended period, and limited the amount of steam generator leak rate information available to the operators before the event.
The AIT determined that the number and duration of the equipment problems reflected weaknesses in engineering, corrective action processes, and operational support at the station. The licensee's response to a number of the equipment problems identified during the event reflected an acceptance of "working around" the problem rather than fixing it.
Emergency Response
The ERO took the necessary steps to ensure the protection of public health and safety. The operators properly classified the event, and the licensee implemented a thorough peer review of the emergency response to this event. The AIT identified several emergency plan and implementing procedure problems similar to those identified by the licensee's peer review team, including the following:
The licensee developed and was in the process of implementing an emergency response improvement plan before the event.
This information notice requires no specific action or written response. However, recipients are reminded that they are required to consider industry-wide operating experience (including NRC information notices) when practical when setting goals and performing periodic evaluations under Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," of Part 50 of Title 10 of the Code of Federal Regulations. If you have any questions about the information in this notice, please contact the one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
| | /RA by John Tappert Acting For/ Ledyard B. Marsh, Chief
Events Assessment, Generic Communications and Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation |
| Technical contacts: | Eric Benner, NRR
301-415-1171
E-mail: ejb1@nrc.gov | Lawrence Doerflein, Region I
610-337-5378
E-mail: ltd@nrc.gov
|
| | Peter Eselgroth, Region I
610-337-5234
E-mail: pwe@nrc.gov | Raymond Lorson, Region I
603-474-3589
E-mail: rkl@nrc.gov |
| Attachment: | List of Recently Issued NRC Information Notices |
(ADAMS Accession Number ML003726124)
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