05000277/LER-2011-001, For Peach Bottom Unit 2 High Pressure Coolant System Inoperable Due to Leaking Cooling Water Header Relief Valve
| ML11131A049 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/06/2011 |
| From: | Stathes G Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 11-001-00 | |
| Download: ML11131A049 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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 |
| 2772011001R00 - NRC Website | |
text
Exelon.
Exelon Nuclear www.exeloncorp.com Nuclear Peach Bottom Atomic Power Station 1848 Lay Rd.
Delta, PA 17314 10CFR 50.73 May 6, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station (PBAPS) Unit 2 Facility Operating License No. DPR-44 NRC Docket Nos. 50-277
Subject:
Licensee Event Report (LER) 2-11-01 Enclosed is a Licensee Event Report concerning a condition involving an inoperability of the Unit 2 High Pressure Coolant Injection (HPCI) system. In accordance with NEI 99-04, the regulatory commitment contained in this correspondence is to restore compliance with the regulations. The specific methods that are planned to restore and maintain compliance are discussed in the LER. If you have any questions or require additional information, please do not hesitate to contact us.
Sincerely, Garey L. Stathes Plant Manager Peach Bottom Atomic Power Station GLS/djf/I R 1188457 / 1195100 / 1195115 Attachment cc:
US NRC, Administrator, Region I US NRC, Senior Resident Inspector R. R. Janati, Commonwealth of Pennsylvania S. Grey, State of Maryland P. Steinhauer, PSE&G, Financial Controls and Co-owner Affairs INPO Records Center CCN: 11-39
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 digits/characters for each block) information collection.
- 3. PAGE Peach Bottom Atomic Power Station (PBAPS) Unit 2 05000277 1 OF 5
- 4. TITLE High Pressure Coolant Injection System Inoperable due to Leaking Cooling Water Header Relief Valve
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MNH DY YA YERSEQUENTA REV MONTH DAY YEAR 05000 MONTH DAY YEAR YEAR SUE RENO.
D FACILITY NAME DOCKET NUMBER 03 16 2011 11 001 00 05 06 2011 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b)
[I 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 1 [1 20.2201(d) 0l 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL [I 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
C1 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[E 50.36(c)(2)
El 50.73(a)(2)(v)(A)
[1 73.71(a)(4) 100%
[E 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B)
[E 73.71 (a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)
ED 50.73(a)(2)(v)(D)
Specify in Abstract below or in Analysis of the Event, continued 1 OCFR 50.73(a)(2)(i)(B) - Condition Prohibited by Technical Specifications - Since there exists evidence that RV-2-23B-066 had likely failed in January 2011, it was determined that HPCI was inoperable for a time period greater than the 14 days allowed by Technical Specification 3.5.1, Emergency Core Cooling Systems (ECCS) - Operating, Required Action C.2. Therefore, this condition is considered to have been prohibited by the Technical Specifications.
There were no actual safety consequences or actual water hammer events associated with this event.
The HPCI system is designed to flood the reactor during design basis events involving loss of cooling to the reactor core. The HPCI system is designed with two suction flow paths: the CST and the Suppression Pool flow paths. Normally, HPCI is aligned in the standby mode to the CST during plant operations. Transfer to the Suppression Pool suction source occurs during design basis events only if there is CST low water level or Suppression Pool high water level.
No credit for the CST is taken in the plant safety analyses.
The RV-2-23B-066 is the relief valve for the 2" HPCI cooling water header. The cooling water header tap comes off the HPCI booster pump discharge and provides regulated cooling flow to HPCI system components (e.g., HPCI lube oil cooler). The leakage through the seat of RV 23B-066, during standby conditions, was approximately 7 gpm of water into the floor drain and provided input into the Unit 2 Reactor Building sump.
During design basis events, HPCI normally starts first with its suction source aligned to the CST. In this configuration, there would have been no voiding concerns on the HPCI discharge piping due to the much higher elevation of the CST suction source versus the water elevation in the Suppression Pool. Although leakage would have occurred through the RV-2-23B-066 seat, there would not have been voiding in the HPCI discharge piping. This was validated by pump, valve and flow surveillances that were completed using the CST as the suction source.
Therefore, for design basis events with HPCI aligned from the normal CST suction source, there would be no significant impact as a result of the RV-2-23B-066 condition. However, for certain design events (e.g. loss-of-offsite-power occurrences, station blackouts, anticipated transient without scram), HPCI may be required to restart for these events. If HPCI restarted while aligned to the CST, there would be no HPCI performance impact as a result of the RV 23B-066 condition. If the suction source had been transferred to the Suppression Pool, then the HPCI discharge piping could be drained of water to the Reactor Building sump while the system is in standby and not running.
If the system was required to restart, then a water hammer condition could exist and the ability for HPCI to perform its intended design function for restart could not be assured due to potential loss of piping integrity.
Analysis of the Event, continued HPCI was determined to be inoperable from approximately 1/20/11 to 3/18/11 when HPCI was declared operable by Operations personnel following the replacement of RV-2-23B-066.
Subsequent investigation determined that RV-2-23B-066 had been recently replaced as a preventive maintenance task performed on 1/19/11.
Based on trending performed on water inputs into the Unit 2 Reactor Building sump, it was identified on 1/25/11 that there had been an increasing trend of water input into the sump that began on approximately 1/20/11. Subsequent to the HPCI inoperability identified on 3/16/11, it was determined that the HPCI RV-2-23B-066 likely began leaking on 1/20/11 and was contributing to the rise in the Reactor Building sump pump-out rate. During this time period, the Reactor Core Isolation Cooling (RCIC) system and the Automatic Depressurization System (ADS) were both operable to support high pressure cooling requirements for design basis events.
As a result of the relief valve failure in the open position, an Engineered Safety Feature (ESF) leak had been occurring.
During design basis events, this leak may have exceeded the allowable 5 gpm assumption in the plant safety analysis. Although this leakage exceeded the ESF leakage requirements assumed in the Alternate Source Term (AST) radiological analysis, an engineering evaluation determined that off-site doses, for design basis events, would remain below the limits required by 10CFR 50.67.
Cause of the Event
The cause of the inoperability of HPCI was due to a leaking relief valve on the HPCI cooling water header. Once identified as leaking on 3/17/11, the relief valve was removed, bench-tested and determined to not be functioning properly.
Additional laboratory testing was performed on the relief valve. Although the cause could not be definitively determined, the cause of the unexpected failure could be the introduction of fine particulate, such as iron oxide, between the disc outside diameter and nozzle bore. The diametrical clearance between the disc outside diameter and nozzle bore measured 0.007". While no particulate was observed on the disc or nozzle, scoring and scuff marks on the disc outside diameter and nozzle bore suggest some interference might have been present.
The 1" relief valve is supplied by Anderson-Greenwood-Crosby (model no. 1 X 1.5 JMB-C-E).
Corrective Actions
The HPCI cooling water header relief valve (RV-2-23B-066) was replaced. Appropriate testing ensured that there were no leaks and the HPCI system was returned to an operable status on 3/18/11.
Additional corrective actions are being pursued in accordance with the corrective action program.
Previous Similar Occurrences There were no previous LERs identified involving HPCI inoperabilities caused by a failure of the HPCI cooling water header relief valve.