IR 05000443/2007007
| ML071730068 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/21/2007 |
| From: | Rogge J Engineering Region 1 Branch 3 |
| To: | Pierre G Florida Power & Light Energy Seabrook |
| References | |
| IR-07-007 | |
| Download: ML071730068 (13) | |
Text
June 21, 2007
SUBJECT:
SEABROOK GENERATING STATION - NRC SUPPLEMENTAL INSPECTION REPORT 05000443/2007007
Dear Mr. St. Pierre:
On May 17, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed a supplemental inspection at the Florida Power and Light Corporations (FPLs) Seabrook Generating Station Unit 1. The enclosed report documents the inspection results that were discussed on May 17, 2007, with Messrs. T. Jones and M. Kiley and other members of your staff.
The NRC performed this supplemental inspection to assess your evaluation of emergency diesel generator (EDG) failures that resulted in the EDG Mitigating System Performance Indicator (MSPI) crossing the Green (very low safety significance) to White (low to moderate safety significance) threshold for the third quarter of 2006. The supplemental inspection was conducted to determine if the root and contributing causes of the failures were understood, to assess the extent of condition review, and to determine if the corrective actions were sufficient to address the identified causes and to prevent recurrence. The inspection was conducted in accordance with Inspection Procedure 95001, Inspection for One or Two White Inputs in a Strategic Performance Area, and examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
Based on the results of this inspection, we concluded that you have adequately evaluated the causes of the failures and the associated performance deficiencies and have identified and implemented appropriate corrective actions. No findings of significance were identified. Given your acceptable performance in addressing the causes of the EDG failures, and the PI returning to Green status in the fourth quarter of 2006, this issue will no longer be considered in assessing Unit 1 plant performance in accordance with the guidance in Inspection Manual Chapter (IMC) 0305, Operating Reactor Assessment Program.
G. S In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (The Public Electronic Reading Room).
Sincerely,
/RA/
John F. Rogge, Chief Engineering Branch 3 Division of Reactor Safety Docket No.
50-443 License No.
NPF-86 Enclosure:
Inspection Report 05000443/2007007 w/Attachment: Supplemental Information
SUMMARY OF FINDINGS
IR 05000443/2007007; 05/14/2007 - 05/17/2007; Seabrook Station, Unit 1; Supplemental
Inspection; IP 95001, Inspection for One or Two White Inputs in a Strategic Performance Area.
The inspection was conducted by a regional inspector. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
Cornerstone: Mitigating Systems
The NRC performed this supplemental inspection in accordance with Inspection Procedure 95001 to assess the licensees evaluation of EDG failures that resulted in the MSPI Index for Emergency Alternating Current Power System crossing the Green to White threshold. Five failures that occurred in the three year period between 2004 and 2006 caused the PI status change in August 2006. Two of the EDG failures were caused by jacket water cooling instrumentation problems and three failures involved over voltage events caused by voltage regulator malfunctions.
FPLs evaluation of the issues included the performance of apparent cause evaluations for the two events associated with jacket water cooling instrumentation failures and a full root cause analysis to identify the root and contributing causes associated with the three over voltage events. The NRC also conducted a special inspection in response to EDG failures that included the final over voltage event in August, 2006. The results of that inspection are documented in inspection report 05000443/2006016.
Based on the results of this inspection, the inspector concluded that FPL completed thorough evaluations of the EDG events, including associated performance deficiencies and implemented appropriate corrective actions to address the related causes. Given FPLs acceptable performance in addressing the EDG failures, and the PI returning to Green status in the fourth quarter of 2006, this issue will no longer be considered in assessing Seabrook plant performance in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program.
REPORT DETAILS
INSPECTION SCOPE
The U.S. Nuclear Regulatory Commission (NRC) performed this supplemental inspection to assess FPLs evaluation of the root and contributing causes for five EDG failures that occurred within the past three years and resulted in the MSPI for Emergency Alternating Current Power System crossing the Green/White threshold.
The inspector performed a walk-down of the affected EDGs, interviewed selected FPL staff, and reviewed documents pertaining to the cause evaluations and corrective actions for the events. The inspector also reviewed the corrective actions to ensure the actions addressed both the root and contributing causes for the identified performance deficiencies.
EVALUATION OF INSPECTION REQUIREMENTS 02.01 Problem Identification a.
Determination of who identified the issue and under what conditions.
Each of the five failures were self-revealing and occurred during the performance of routine surveillance tests or during post-maintenance testing.
b. Determination of how long the issue existed and prior opportunities for identification
.
FPLs evaluations determined that the two issues involving jacket water cooling system instrumentation failures were caused, in part, by maintenance work activities and were promptly discovered during the subsequent post-maintenance surveillance test. Those events had not presented the licensee with prior opportunities for identification.
However, the over voltage events involved long standing hardware deficiencies within the voltage regulators. The root cause evaluation for these events recognized that there had been missed opportunities for identification and correction of the cause of the failures. Corrective actions were implemented to address the failure to more promptly identify and correct the causes of the over voltage events.
c.
Determination of the plant-specific risk consequences and compliance concerns associated with the issue.
The failures of the jacket water cooling instrumentation occurred, and were identified and corrected, during periods when the EDGs were out-of-service for scheduled maintenance. The root cause evaluation of the over voltage events included an assessment of the risk consequences for the voltage regulator failures and determined the increase in core damage frequency (CDF) was approximately 6.88E-7. This assessment was consistent with an independent risk evaluation performed by a Region I senior reactor analyst.
The event involving the over voltage condition in August 2006 was determined to be reportable to the NRC as required by 10 CFR 50.73 and was reported in Licensee Event Report (LER) 2006-006-001.
02.02 Root Cause and Extent of Condition Evaluation a.
Evaluation of methods used to identify root cause and contributing causes.
The inspector found that the licensee utilized systematic methods to determine the root and contributing causes for each of the EDG failures. The apparent cause evaluations for the jacket water cooling failures used the 5 Whys root cause analysis method while the root cause evaluation for the over voltage events utilized both the Events and Causal Factors and Barrier Analysis techniques. Additionally, identification of the causes for the over voltage events involved significant in plant troubleshooting, circuit analysis and simulation, laboratory mock-up testing and individual component failure analysis methods.
b.
Level of detail of the root cause evaluation.
The level of detail of the cause evaluations were found to be appropriate and commensurate with the significance of the event as well as the complexity of the root and contributing causes. For example, the over voltage event investigation included an evaluation of why station processes and programs had not been effective in identifying and correcting the cause of the events when addressing previous similar failures.
c.
Consideration of prior occurrences of the problem and knowledge of prior operating experience.
For each event the licensee cause evaluation did consider prior occurrences and available operating experience. In particular, the evaluation identified that a number of prior over voltage events had occurred and that the corrective actions for those events had not been effective.
d.
Consideration of potential common causes and extent of condition of the problem.
Each of the cause investigations were thorough in assessing the extent of condition and extent of cause for the event. For causes that involved equipment failures, similar other components within the affected system as well as similar components in other systems were assessed and appropriate actions were taken to prevent common cause failures.
As a result of the over voltage event evaluation, the extent of cause evaluation included actions to assess other systems which had experienced repeat failures to determine if additional actions were necessary. These actions identified a similar situation involving repeat component failures and resulted in the initiation of a full root cause investigation to identify the root cause of diode failures in safety-related circuit breakers.
e.
Consideration of safety culture components as described in Inspection Manual Chapter 0305.
The inspector found that FPLs evaluation included appropriate consideration of safety culture components and has implemented corrective actions for those items that were determined to have been contributing causes. These components included the trending and evaluation elements of the corrective action program.
02.03 Corrective Actions a.
Appropriateness of corrective actions.
The inspector found that FPL established appropriate corrective action plans for each of the events. The plans included immediate actions to restore the affected EDG to an operable condition and to address extent of condition issues. The plans also included longer term items to address items such as improvements to station program and processes, training of personnel, hardware upgrades and to assess the need for design changes.
b.
Prioritization of corrective actions.
The inspector reviewed the corrective action plan to determine the status of each corrective action item and found that most of the items had been already completed and that the remaining items had been appropriately prioritized.
c.
Establishment of schedule for implementing and completing the corrective actions.
The inspector confirmed that the remaining open corrective actions have been assigned acceptable completion dates. Actions that were not already complete do not impact the EDG reliability or availability and have been scheduled to be accomplished at the next opportunity.
d.
Establishment of quantitative or qualitative measures of success for determining the effectiveness of the corrective actions to prevent recurrence.
The inspector reviewed the corrective action plan and verified that corrective action CR 06-010146-05 has been established to verify that 1) corrective actions have been completed, 2) no additional similar conditions adverse to quality had occurred, and 3) that the corrective actions to prevent recurrence have not resulted in any new conditions adverse to quality.
MANAGEMENT MEETINGS
Exit Meeting Summary
The results of this inspection were discussed with Messrs. T. Jones, Vice President -
Support, and M. Kiley, Plant General Manager, and other members of the Seabrook staff at the conclusion of this inspection on May 17, 2007. Following the exit meeting, a Regulatory Performance Meeting was conducted, by conference call, in accordance with Inspection Manual Chapter 0305, Operating Reactor Assessment Program, and focused on the performance deficiencies associated with this issue and corrective actions to prevent recurrence. Mr. T. Jones (FPL) and Mr. John Rogge (NRC)participated in the meeting by telephone. No proprietary information was discussed or has been included in this report.
ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
R. Arn
System Engineer
J. Ball
MSPI Coordinator
P. Brangiel
I&C System Engineer
R. Jamison
Design Engineer
T. Jones
Vice President-Support
M. Kiley
Plant General Manager
G. Kotowski
Electrical Design Supervisor
M. OKeefe
Regulatory Compliance Supervisor
M. Makowicz Plant Engineering Manager
T. Manning
System Engineer
V. Roberts
Sr. Nuclear Analyst
D. Wilson
Engineering Support Supervisor
NRC Personnel
- W. Raymond, Senior Resident Inspector, Seabrook Station
- S. Shaffer, Resident Inspector, Seabrook Station
ITEMS OPENED, CLOSED, AND DISCUSSED
None