05000382/LER-2004-001, Regarding Failure to Provide Backup Over-Current Protection Due to Personnel Error
| ML040840529 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/22/2004 |
| From: | Sen G Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| W3F1-2004-0012 LER 04-001-00 | |
| Download: ML040840529 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3822004001R00 - NRC Website | |
text
Entergy Operations, Inc.
nfoers 17265 River Road n t Killona, LA 70066
~
Tel 504 739 6650 10CFR50.73 (a)(2)(i)(B)
W3Fl -2004-0012 March 22, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Licensee Event Report 2004-001-00 Gentlemen:
Attached is Licensee Event Report (LER) 2004-001-00 for Waterford Steam Electric Station Unit 3. This report provides details of an event where it was determined that two electrical AC circuits, which go through containment electrical penetrations, did not have backup over-current protection as required by the Technical Specification.
This condition is being reported pursuant to 10CFR50.73 (a)(2)(i)(B) as a condition prohibited by the Technical Specification. There are no commitments contained in this submittal. If you have any questions, please contact Michael E. Mason at (504) 739-6673.
Very truly yours, G. Sen Manager, Licensing GS/MEM/cbh
/lUIc
- l III MNI IL
W3F1 -2004-0012 Page 2 cc:
Mr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop 0-07 D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P.O. Box 4312 Baton Rouge, LA 70821-4312 R.K. West, lerevents@inpo.org - INPO Records Center
Abstract
On January 21, 2004, it was determined that two 120 volts electrical AC circuits, each going through separate containment electrical penetrations, did not have in-line backup over-current protection as required by the Technical Specification. The circuits in question power the plant's hydrogen analyzer panels and position indication for the hydrogen analyzer containment isolation valves. Upon this discovery, the plant followed the required action statement for the Technical Specification, 3.8.4.1, by lifting the circuit leads for the inside containment isolation valve position indication only, declaring the affected position indication and circuit inoperable and verifying every 7 days that the circuits were disconnected. The hydrogen analyzer remained capable of performing its' intended safety function.
Backup over-current protection was added as the corrective measure. The cause of this event was human error in that personnel did not perform adequate peer and self checking when modifying the affected circuits in 1988 during installation of a station modification package. A contributing cause was determined to be that the control wiring diagrams for the Containment Hydrogen Analyzers did not clearly distinguish the power supply for the containment penetration isolation valves position indication. The lack of in-line backup over-current protection for the above circuits did not result in damage to the containment penetrations. Accordingly, this event did not compromise the health and safety of the public.
NRC FORM 366 (7-2001)U.S. NUCLEAR REGULATORY COMMISSION
.7-2001))
LICENSEE EVENT REPORT (LER)
FACILITY NAME ( 1 DOCKET (2 j LER NUMBER(
PAGE (36 YEAR SEQU ENTIAL REVIS2ON NUMBER NUMBER Waterford Steam Electric Station, Unit 3 05000-382 2004 001 00 2OF 5 TEXT (If more space is rqwred, use additional copies of NRC Form 366A) (17)
REPORTABLE OCCURRENCE On January 21, 2004, it was determined that two electrical AC circuits, which go through containment electrical penetrations, did not have backup over-current protection as required by the Technical Specification. The applicable Technical Specification action statement, 3.8.4.1 a, requires in part that with one or more of the containment penetration conductor over-current devices inoperable:
- 1. Restore the protective device(s) to OPERABLE status or de-energize the circuit(s) by tripping, racking out, or removing the alternate device or racking out or removing the inoperable device within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
- 2. Declare the affected system or component inoperable, and
- 3. Verify at least once per 7 days thereafter the alternate device is tripped, racked out, or removed, or the device is racked out or removed.
For the event described in this report, the back-up in-line over-current protective devices with the hydrogen analyzer panels were inadvertently removed in 1988 during the modification of the two trains of Containment Hydrogen Analyzers and remained in that condition until January 2004.
Therefore, this event is being reported per 10CFR50.73 (a)(2)(i)(B), as an operation or condition prohibited by the plant's Technical Specifications, as the action statement was not performed within the allowed outage time.
INITIAL CONDITIONS Prior to the discovery of this event, the plant was operating at 100% in Mode 1. There were no procedures being implemented specific to this event. There were no Technical Specification Limiting Conditions of Operation specific to this event in effect. There was no major equipment out of service specific to this event out of service.
EVENT DESCRIPTION
In January 2004, during a review of Waterford 3 work packages associated with the hydrogen analyzers, it was discovered that circuit breakers [ED] feeding the Containment Hydrogen Analyzer Analysis Units [IK] A and B were not listed in the Technical Requirements Manual as protecting the two separate containment [NH] electrical penetrations as required. A further review by Waterford 3's Engineering personnel determined the related circuits were missing backup in-line over-current protection as required by Technical Specification 3.8.4.1.
Background
Containment Hydrogen Analyzers A and B were installed in the plant during initial construction.
The power feed to the Hydrogen Analyzers was provided using double breaker combination of lNRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (7-2001))
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET(2 LER NUMBER (6)
PAGE 3)
YEAR SEQUENTAL REVISION NUMBER NUMBER Waterford Steam Electric Station, Unit 3 05000-382 2004 001 00 3 OF 5 TEXT (If more space isrequired, use additional copies of NRC Form 366A) (17) primary circuit breakers and backup circuit breakers.
In 1988, both trains of the Containment Hydrogen Analyzers were modified by a station modification package (SMP). The SMP split the power feed for each Containment Hydrogen Analyzer into two different circuits. One circuit (circuit 7) supplied the new containment hydrogen analyzer control panel and containment isolation valve position indication protected only by one circuit breaker. The other circuit (circuit 43) supplied the containment hydrogen analyzer sample isolation valves and containment isolation valves protected by one circuit breaker and a fuse.
However, it was determined that the power needs of the containment hydrogen analyzer sample isolation valves and containment isolation valves were not met by the electrical circuit selected in the SMP, therefore another circuit (circuit 32) was selected, but this circuit did not provide backup over-current protection.
In 1995, the circuit (circuit 32) supplying the containment hydrogen analyzer sample isolation valves and containment isolation valves was identified to be missing backup over-current protection. This condition was reported in LER 1995-006-00, and a fuse was added to the circuit.
In 2004, the circuit (circuit 7) that supplied the new containment hydrogen analyzer control panel and containment isolation valve position indication by a circuit breaker was determined to not have backup over-current protection as required by Technical Specification 3.8.4.1.
CAUSAL FACTORS The cause for not having backup over-current protection for position indication for the containment isolation valves was personnel error during the preparation and reviews of the SMP. A contributing cause was inadequate documentation, in that control wiring diagrams for the Containment Hydrogen Analyzers did not clearly distinguish the power supply for the position indication for containment isolation valves.
CORRECTIVE ACTIONS
Fuses were added to the circuits as backup over-current protection for containment isolation valves position indication in February 2004. The circuit breakers for the containment hydrogen analyzer control panel and containment isolation valves position indication installed during the fuse addition to the circuit were tested and demonstrated operable. The circuit breakers and fuses were added to the TRM containment electrical penetration table, which contains the list of containment electrical penetrations to be tested in accordance with Technical Specification Surveillance Requirement 4.8.4.1.
Other actions initiated include: generate tasks to test the circuit breakers in accordance with the Technical Specification requirements, relocate the control circuit test from a preventative maintenance procedure to a Technical Specification Surveillance test procedure, revise the applicable Control Wiring Diagrams, and develop and implement lessons learned in training for engineering personnel.
Additionally, Engineering performed a review of all electrical containment penetration circuits to verifyU.S. NUCLEAR REGULATORY COMMISSION 77-2001))
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER PAGE 3 YEAR SEQUENTIAL REVISION NUMBER NUMBER Waterford Steam Electric Station, Unit 3 1050003821 2004 001 00 4 OF 5 TEXT (If more spaceis required, use additional copes of NRCForm 366A) (17) that all other containment electrical penetrations were protected by both primary and backup over-current devices as required by Technical Specification 3.8.4.1. During this review, no conditions prohibited by Technical Specification 3.8.4.1 were revealed, however, some minor discrepancies were identified with corrective actions pending.
SAFETY SIGNIFICANCE
Backup over-current protection is required for containment electrical penetrations by Technical Specification. The backup over-current protection ensures that containment penetration physical integrity is not impaired in the event of an electrical fault inside containment and the failure of one electrical device to interrupt and isolate the fault. During the period when no backup over-current protection existed for the containment electrical penetration circuits, the primary breakers were being satisfactorily tested to an equivalent test methodology used for testing circuit breakers within the allowed frequency of the Technical Specification requirements. Therefore, if a faulted condition had occurred for the circuits in question, the primary breakers would have opened to protect the containment electrical penetrations.
The risk assessment for this condition addressed two aspects: containment isolation function and potential change in Core Damage Frequency (CDF). With a circuit fault and loss of the valve position indication, the containment isolation valves would still have been capable of performing its' safety function of closure since they were powered from a different circuit and are fail closed solenoid valves (an exception is that one of the containment isolation valves is a check valve which would seat since normal direction of flow is into containment).
For the worst case scenario, if an electrical fault occurred which failed to open the primary circuit breaker on over-current as designed, then the potential existed for a one inch diameter hole to develop in the containment penetration. However, this would be precluded by various factors (e.g.
primary breakers were capable of tripping on fault condition and operations monitor system(s) performance). In the extremely unlikely event that this failure were to occur in conjunction with a large break loss of coolant accident, the leak rate from containment would be greater than the assumed leak rate in the loss of coolant accident offsite and control room radiological dose consequences analysis.
When combining the probability of core damage frequency in combination with the electrical fault condition on the penetration, the risk impact was determined to be negligible. The electrical fault condition on the penetration had no impact on Core Damage Frequency. The electrical fault condition on the penetration had a negligible impact on Large Early Release Frequency (LERF) in that the one inch diameter hole is bounded by the two inch diameter hole in containment assumed in the LERF. The frequency of one of the protective breakers failing concurrent with a core damage event given an electrical fault condition on the penetration would be 4E-9 per year.
This event is not a Safety System Functional Failure (SSFF).U.S. NUCLEAR REGULATORY COMMISSION
-(7-2001))
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER Waterford Steam Electric Station, Unit 3 05000-382 2004 001 00 5 OF 5 TEXT (If more space is required, use additonal copies of NRC Foifn 366A) (17)
SIMILAR EVENTS
A similar event was reported to the NRC in LER 95-006-00, that found backup over-current protection devices missing for the Containment Hydrogen Analyzer containment isolation and sample valves. The cause for not having backup over-current protection for the Containment Hydrogen Analyzer containment isolation and sample valves was personnel error.
ADDITIONAL INFORMATION
Energy Industry Identification System (EIIS) codes are identified in the text within brackets [.