05000247/LER-2002-001, Re While Preparing for Maintenance Activities Personnel Error Results in Two Emergency Diesel Generators Inoperable
| ML020720022 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  (DPR-026) |
| Issue date: | 03/01/2002 |
| From: | Dacimo F Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-02-023 LER-02-001-00 | |
| Download: ML020720022 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2472002001R00 - NRC Website | |
text
Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
E g
Indian Point Energy Center Entergy 295 Broadway, Suite 1 P.O. Box 249 Buchanan. NY 10511-0249 March 1, 2002 Re: Indian Point Unit No. 2 Docket No. 50-247 LER 2002-001-00 NL-02-023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-Pl-17 Washington, DC 20555-0001
Dear Sir:
The attached Licensee Event Report 2002-001-00 is hereby submitted in accordance with the requirements of 10 CFR 50.73.
There are no commitments contained in this letter.
S
- ely, Fred Dacimo Vice President -
Operations Indian Point 2 Attachment cc:
Mr. Hubert J. Miller Regional Administrator -
Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Patrick D. Milano, Senior Project Manager Project Directorate I Division of Licensing Project Management U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington, DC 20555 Senior Resident Inspector U.S. Nuclear Regulatory Commission PO Box 38 Buchanan, NY 10511
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2001 (6-1998)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Indian Point, Unit 2 05000247 1 OF 5
While Preparing for Maintenance Activities Personnel Error Results in Two Emergency Diesel Generators Inoperable EVENT DA E (5)
LER NUMBER 16)
OTHER FACILITIES INVOLVED (8)
F iACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION05 0
MONTH DAY YEAR YEAR I
NUMBER MONTH DAY YEAR A
0500C0 llFACILITY NAME lDOCKET NUMBERl 01 06 2002 2002
- - 001-00 03 01 2002 l
05000 OPERATING 1 1-THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)
MODE (9) l N
I 20.2201(b) l 20.2203(a)(2)(v) l 50.73(a)(2)(i) 50.73(a)(2)(viii POWER 1 1 l 20.2203(a)(1) l 20.2203(a)(3)(i)
X 50.73(a)(2)(ii) 50.73(a)(2)(x)
LEVEL (10) I 100 ll 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) l 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)
OTHER-20.2203(a)(2)(iii) 50.36(c)(1)
X 50.73(a)(2)(v)
- - Specify in Abstract below or
___20_2203(a)(___
50.36(c)(2)
I 150-73(a)(2)(vii) in NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
Richard Louie, Licensing Engineer 1
(914) 734-5678 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
.!1 REPORTABLE REPORTABLE
CAUSE
- - SYSTEM COMPONENT MANUFACTURER TO EPIX l
CAUSE
SYSTEM COMPONENT MANUFACTURER TO EPIX l
lIPPI FMFNTAI RFPQRT FXPFfTFnia Jt8l XPCE MONTH l-DAY l
YEAR EXPECTED leYES l
l I
SUBMISSION l
l l
! If y, es, complete EXPECTED SUBMISSION DATE).l X INO DATE (15) l______
l l
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On January 6, 2002, at approximately 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, while performing an independent verification of tagging operations prior to planned maintenance activities on 23 Emergency Diesel Generator (EDG)[EIIS:EK], an individual performing the verification mistakenly actuated the overspeed trip lever for 21EDG.
This condition resulted in the unavailability of both 21 and 23EDGs, and placed the plant in a shutdown action statement per Technical Specification 3.0.1.
The individual immediately recognized the error, and initiated corrective actions to restore 21EDG to operable condition.
21EDG was returned to service in less than five minutes, and the action statement of Technical Specification 3.0.1 was subsequently exited.
At the time, the plant was at 100 percent power, and normal offsite power feeders were available.
No injuries to plant personnel, damage to any equipment, or adverse safety implications to the public occurred as a result of this event.
The EDG system is designed such that any two of the three dedicated power trains are capable of supplying the minimum engineered safety features loads, assuming the single failure of one power train or EDG.
This report is submitted pursuant to 10 CFR 50.73(a)(2)(v) as a condition that alone could have prevented the fulfillment of the safety function of a structure or system needed to mitigate the consequences of an accident, and 10 CFR 50.73(a)(2)(ii) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
NRC FORM 366 (6-1998)U.S. NUCLEAR REGULATORY COMMISSION (6-1 998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION DOCKET FACILITY NAME (1)
NUMBER (2)
LER NUMBER (6) l PAGE (3)
SEQUENT l REVISI IAL ON YEAR NUMBER NIIMRF Indian Point, Unit 2 05000247 2002
- - 001-00 12 OF 5
TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION
Westinghouse 4-Loop Pressurized Water Reactor EVENT IDENTIFICATION While Preparing for Maintenance Activities Personnel Error Results in Two Emergency Diesel Generators Inoperable EVENT DATE January 6, 2002 REFERENCES Condition Reporting System Number: 200200192 PAST SIMILAR EVENTS LER 2001-002, 1999-018, 1997-016
EVENT DESCRIPTION
On January 6, 2002, at approximately 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, while performing an independent verification of tagging operations prior to planned maintenance activities on 23 Emergency Diesel Generator (EDG), an individual performing the verification mistakenly actuated the overspeed trip lever for 21EDG.
This condition resulted in the unavailability of both 21 and 23EDGs, and placed the plant in a shutdown action statement per Technical Specification 3.0.1.
The individual immediately recognized the error, and initiated corrective actions to restore 21EDG to operable condition.
Control room personnel were immediately aware of this condition having received a "Diesel Generator Trouble" alarm actuation.
The overspeed trip lever was reset, the alarm was cleared, and 21EDG was returned to operable status in less than five minutes.
The action statement of Technical Specification 3.0.1 was subsequently exited.
23EDG remained inoperable as initially intended for maintenance activities.
With 23EDG unavailable the limitations of Technical Specification 3.7.B.l.a remained in effect.
At the time, the plant was at 100 percent power, and normal offsite power feeders were available.
No injuries to plant personnel, damage to any equipment, or adverse safety implications to the public occurred as a result of this event.U.S. NUCLEAR REGULATORY COMMISSION (6-1 998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION DOCKET FACILITY NAME (1)
NUMBER (2)
LER NUMBER (6)
PAGE (3) l SEQUENT I REVISI IAL ION YEAR I NUMBER NIjMR Indian Point, Unit 2 05000247 2002
- - 001 00 3 OF 5
TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
EVENT DESCRIPTION (Continued)
The EDG system is designed such that any two of the three dedicated power trains are capable of supplying the minimum engineered safety features loads, assuming the single failure of one power train or EDG.
This report is submitted pursuant to 10 CFR 50.73(a)(2)(v) as a condition that alone could have prevented the fulfillment of the safety function of a structure or system needed to mitigate the consequences of an accident, and 10 CFR 50.73(a)(2)(ii) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
At the time, it was believed that no notifications pursuant to 10 CFR 50.72 were necessary, because although two EDGs were unavailable, one EDG power train would have been adequate to provide power for a safe and orderly plant shutdown in the event of a loss-of-offsite electrical power.
Subsequent review determined that a single EDG power train would not have been adequate under design basis accident conditions.
EVENT ANALYSIS
The expectations and requirements for conducting independent verifications are identified in Operations Administrative Directive (OAD)-6, "Equipment Status Control.'
These requirements were recently revised such that those individuals performing independent verifications could not participate in the specific equipment tagout activities.
Furthermore, concurrent verification while tagging out equipment is not acceptable.
Thus, it is expected that independent verifications be conducted upon the completion of the specific tagout activities.
As part of the tagout activities for 23EDG, a Nuclear Plant Operator (NPO) was directed to disable the auto start capability of the EDG by setting its overspeed trip lever to the tripped position.
This was performed by the NPO as expected, and confirmed via illumination of the local control panel indication light.
The position of the EDG overspeed reset lever is also an observable indication of the tripped status of the engine.
Upon completion of the aforementioned tagout activities, the individual performing the independent verification wanted to confirm the status of 23EDG and proceeded to pull the overspeed lever; however, the overspeed trip lever on 21EDG was mistakenly actuated.
The cause for this event is human performance related has been attributed to:
- 1. Inadequate implementation of self-checking techniques such as "STAR."
- 2. Misunderstanding of the appropriate method to confirm and verify component positions during independent verifications.
- 3. Lack of component labeling sufficient to create a mental barrier preventing crossing over to active systems or trains.
With the exception of 23EDG having been declared inoperable, there were no other structures, systems, or components which were inoperable at the start of the event and that contributed to the event.U.S. NUCLEAR REGULATORY COMMISSION (6-1 998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION DOCKET FACILITY NAME (1)
NUMBER (2)
LER NUMBER (6)
PAGE (3) l SEQUENT REVISI IAL ON YEAR NUMBER NlIMR Indian Point, Unit 2 05000247 2002
- - 001 00 4 OF 5
TEXT (if more space is required, use additional copies of NRC Form 366A) (17) 10 CFR 50.72 Notification This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D) as, "Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident."
However, at the time, it was believed that no notifications pursuant to 10 CFR 50.72 were necessary, because although two EDGs were unavailable, one EDG power train would have been adequate to provide power for a safe and orderly plant shutdown in the event of a loss-of-offsite electrical power.
Upon further review it was determined that a single EDG power train would not have been adequate under design basis accident conditions.
In hindsight, an eight-hour report notification pursuant to 10 CFR 50.72(b)(3)(v) was warranted.
EVENT SAFETY SIGNIFICANCE This event was potentially significant because it resulted in the unavailability of two of the three EDGs.
The design basis of the EDG system assures that any two of the three power trains and dedicated EDGs are capable of supplying the minimum engineered safety feature loads, assuming a single failure of one power train or EDG.
However, because the actual time period that both EDGs were unavailable was no greater than five minutes, this event had minimal actual safety implications.
The unavailability of the two EDGs for the brief period of time prior to restoration of 21EDG resulted in an incremental core damage probability (ICDP) of approximately 6E-9.
CORRECTIVE ACTIONS
Immediate corrective actions taken were to reset 21EDG's overspeed trip lever, clear the trouble alarm, and return 21EDG to operable status.
As a result of this event, the following corrective actions have been, or will be implemented.
- 1.
Each Shift Manager will review this event, its safety significance, and re-enforce the need to utilize self-checking techniques with their operating crews.
(Completed)
- 2. Upgrade labeling of the EDGs.
(Due Date: May 15, 2002)
- 3. Evaluate the adequacy of labeling on other similarly located equipment.
(Due Date: May 15, 2002)
- 4. Revise procedures as necessary to require the use of barrier tape as an aid to restrict access to operable equipment whenever work activities involve closely located similar equipment.
(Due Date: May 15, 2002)U.S. NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION DOCKET FACILITY NAME (1)
NUMBER (2)
LER NUMBER (6)
PAGE (3) l SEQUENT REVISI YEAR NUMBER NU IMR Indian Point, Unit 2 05000247 2002
- - 001 00 5 OF 5
TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
PREVIOUS OCCURRENCES
A review of previous occurrences that involved the same underlying concern or reason (i.e., human performance, component operation related) as this event was performed.
Three events were identified, and reported to the NRC in the following LERs:
LER 2001-002-00: This LER reported that on February 14, 2001 during surveillance testing of 480 volt Bus 3A undervoltage relays, a technician performing the test was required to hold a switch in the bypass position while two test buttons were being pressed.
That particular switch was not held in the "test" position as required by procedure, resulting in a trip of the breaker feeding the bus under test and the autostart of the EDGs.
LER 1999-018-00: This LER reported that on October 6, 1999 during the implementation of a temporary facility change, maintenance personnel incorrectly disconnected wires from the load side of a fuse instead of the line side.
This rendered control of the Bus 2A supply and load breakers inoperable for approximately one minute.
During this interval, Bus 2A degraded voltage actuation channels were inoperable due to lack of 125 volt DC power.
LER 1997-016-00: This LER reported that on July 2, 1997 during maintenance activities, a mechanic working on 480 volt Breaker inadvertently depressed the trip button for Breaker 52/2AT3A.
inadvertently de-energized 480 volt Bus 3A, which actuated the logic automatically starting the EDGs.
the conduct of 52/2AT5A This undervoltage