05000382/LER-2012-003, Regarding Cracked Instrument Line Affects Fire Protection Safe Shutdown Analysis
| ML12157A324 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/04/2012 |
| From: | Jacobs D Entergy Nuclear South |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| W3F1-2012-0044 LER 12-003-00 | |
| Download: ML12157A324 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3822012003R00 - NRC Website | |
text
SEntergy Entergy Nuclear South Entergy Operations, Inc.
17265 River Road Killona, LA 70057-3093 Tel 504-739-6660 Fax 504-739-6678 djacob2@entergy.com Donna Jacobs Vice President - Operations Waterford 3 10 CFR 50.73 W3F1-2012-0044 June 4, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike Rockville, MD 20852
Subject:
Licensee Event Report 2012-003-00 Waterford Steam Electric Station, Unit 3 (Waterford 3)
Docket No. 50-382 License No. NPF-38
Dear Sir or Madam:
Entergy is hereby submitting Licensee Event Report (LER) 2012-003-00 for Waterford Steam Electric Station Unit 3. This report provides details associated with a cracked instrument line which affects the Fire Protection Program analysis.
Based on plant evaluation, it was determined that this condition is reportable under 10 CFR 50.73(a)(2)(ii) requirements.
This report contains no new commitments. Please contact Michael Mason, Licensing Manager (acting), at (504) 739-6673 if you have questions regarding this information.
Sincerely, DJIWH
Attachment:
Licensee Event Report 2012-003-00
W3F1-2012-0044 Page 2 cc:
Mr. Elmo E. Collins, Jr.
Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 E. Lamar Blvd.
Arlington, TX 76011-4511 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 0-07D1 Washington, DC 20555-0001 RidsRgn4MailCenter@nrc.gov Marlone. Davis@nrc.gov Dean.Overland@nrc.gov Kaly.Kalyanam@nrc.gov INPO Records Center lerevents@inpo.org
Attachment to W3F1 -2012-0044 Licensee Event Report 2012-003-00
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/31/2013 (10-2010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
j 3. PAGE Waterford 3 Steam Electric Station 05000 382 1 OF 3
- 4. TITLE Cracked Instrument Line Affects Fire Protection Safe Shutdown Analysis
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE 1
- 8. OTHER FACILITIES INVOLVED MONTH DAY IYEAR YEAR SEQUENTIAL IREV MONTH DY EA FCITYNMDOKTUBR NUMBER NO H
DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 04 04 12012 2012'- 003 - 00 06 04 2012 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[1 20.2201(b)
[] 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
LI 50.73(a)(2)(vii) 1 LI 20.2201(d)
E] 20.2203(a)(3)(ii) 0I 50.73(a)(2)(ii)(A)
[: 50.73(a)(2)(viii)(A)
LI 20.2203(a)(1)
[] 20.2203(a)(4) 0 50.73(a)(2)(ii)(B)
[L 50.73(a)(2)(viii)(B)
__ 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
LI 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL EI 20.2203(a)(2)(ii)
E] 50.36(c)(1)(ii)(A)
[] 50.73(a)(2)(iv)(A)
[] 50.73(a)(2)(x)
LI 20.2203(a)(2)(iii)
[] 50.36(c)(2)
E3 50.73(a)(2)(v)(A)
[] 73.71(a)(4) 100 LI 20.2203(a)(2)(iv)
[] 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
E] 73.71(a)(5)
LI 20.2203(a)(2)(v)
E] 50.73(a)(2)(i)(A)
[: 50.73(a)(2)(v)(C)
[
OTHER Specify in Abstract below or in 0_ 20.2203(a)(2)(vi)
Li-50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
REPORTABLE OCCURRENCE On 04/04/2012, during the conduct of a scheduled surveillance, steam leakage was observed from an instrument line associated with the steam supply to Emergency Feedwater (EFW) [BA] Pump AB [P].
Investigation determined the leakage source as a crack in the instrument line downstream of the normally open instrument root valve. The leakage was isolated by closing the instrument root valve.
Engineering evaluation has determined that EFW Pump AB remained capable of performing its safety function. Due to the steam leak in the EFW Pump AB area, the fire protection safe shutdown manual actions could be affected which requires this to be reported under 10CFR50.73(a)(2)(ii)..
INITIAL CONDITIONS During this time period, Waterford Steam Electric Station Unit 3 (Waterford 3) was operating in Mode 1, stable at or near 100% power. The Emergency Feedwater (EFW) system was aligned for normal operations with no plant protection system actuation signals present.
EVENT DESCRIPTION
While Operations personnel were performing the procedure for the EFW Pump AB operability test as a scheduled surveillance, personnel at the pump discovered that a small steam leak previously thought to be coming from the packing of manual isolation valve MS-414 (EFW Pump AB Turbine Steam Supply MS IPT8340 Root) [P] was actually emanating from a crack in the instrument line connected downstream of MS-414. The condition was reported to Control Room personnel and the decision was made to stop the surveillance. In stopping the pump and returning to standby, the EFW Pump AB turbine steam supply is isolated from Main Steam (MS) system pressure, which resulted in stopping the steam leakage.
Closer examination confirmed that the source of the steam leakage was a crack in the instrument line just downstream of valve MS-414. After consultation with Engineering, Operations supervision directed isolating the cracked instrument line by closing valve MS-414.
Operations and Engineering evaluated the status of EFW Pump AB and determined that, with the instrument line isolated, the pump was capable of performing its safety function and was operable.
The steam leak was originally identified as a packing leak on 9/6/2011.
A subsequent Engineering evaluation which considered the as-found steam leakage and location determined that personnel access to EFW Pump AB area was not assured prior to identifying the cracked instrument line.
This reported condition is entered into the site corrective action program as CR-WF3-2012-1669.
SYSTEM DESIGN The EFW system consists of three pumps which take a common suction from the Condensate Storage Pool, discharge to a common header, and supply water to either of two Steam Generators (SGs) [HX]
through redundant flow control and isolation valves. EFW Pumps A and B are 50% capacity motor driven pumps. EFW Pump AB is a 100% capacity steam turbine driven pump. EFW system controls are located both in the Control Room and in the Remote Shutdown Room. All three EFW pumps
automatically start on Emergency Feedwater Actuation Signal (EFAS) and act, along with flow control and isolation valves, to automatically maintain water level in the non-faulted steam generators [HX] to provide an adequate heat sink for residual heat removal post-trip. The system remains in standby unless actuated in response to a loss of Main Feedwater [SJ] to the SGs.
CAUSAL FACTORS The cause of the instrument line crack is not currently known. The crack location is immediately downstream of the closed manual isolation valve MS-414. Due to the close proximity of the isolating valve and the risk of performing the line removal near the operable EFW Pump AB, the instrument line has not yet been removed for further evaluation.
CORRECTIVE ACTIONS
The following corrective actions are in place:
EFW Pump AB was evaluated as operable with a compensatory measure. The compensatory measure is to maintain isolation valve MS-414 closed.
The condition will remain open within the corrective action program until resolved.
SAFETY SIGNIFICANCE
There is minimal safety significance for this condition because, in evaluating past operability, there was minimal adverse affect to mitigating system functions.
Engineering evaluation has determined that the identified crack at its current size would not have resulted in there being insufficient steam available for EFW Pump AB for the evaluated time to establish shutdown cooling. It also would not have resulted in a harsh environment in the Reactor Auxiliary Building (RAB) -35 ft. el. hallway or exceeding the previously analyzed offsite dose release. Therefore, if there had been an accident, the identified crack would not have resulted in any unacceptable consequences.
As the current evaluation cannot rule out crack growth, further engineering evaluation does conclude that EFW Pump AB maintains sufficient steam supply to perform its safety function for the evaluated time to establish shutdown cooling assuming the instrument line were to part at the crack. Control Room dose and offsite dose release are minimally affected. This condition could result in limiting personnel access to the RAB -35 ft. el. hallway, but would not affect operation of EFW Pump AB.
SIMILAR EVENTS
Corrective action program data for the past three years was searched for similar failures. None were found.
ADDITIONAL INFORMATION
Energy industry identification system (EIIS) codes are identified in the text within brackets [].