05000278/LER-2018-002-01, Reactor Coolant System Pressure Boundary Leakage Resulting in Technical Specification Required Shutdown
| ML19010A037 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/21/2018 |
| From: | Herr M Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CCN: 18-131 LER 2018-002-01 | |
| Download: ML19010A037 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 2782018002R01 - NRC Website | |
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~ Exelon Generation CCN: 18-131 December 21, 2018 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Peach Bottom Atomic Power Station (PBAPS) Unit 3 Renewed Facility Operating License No. DPR-56 NRC Docket No. 50-278 Licensee Event Report (LER) 3-18-002, Revision 1 10CFR 50.73 Enclosed is a revision to LER 3-18-002 to identify a second violation of Technical Specifications (TS) that occurred during a TS required plant shutdown. Revised sections are identified by revision bars in the right margin.
In accordance with NEI 99-04, the regulatory commitment contained in this correspondence is to restore compliance with the regulations.
The specific methods that have been 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 Matt Retzer at 717-456-4351.
Sincerely, (Y)~17L Matthew J. Herr Plant Manager Peach Bottom Atomic Power Station MJH/dnc:1/IR 4175898 Enclosure cc:
US NRC, Administrator, Region I US NRG, Senior Resident Inspector R. R. Janati, Commonwealth of Pennsylvania D. Tancabel, State of Maryland B. Watkins, PSE&G, Financial Controls and Co-Owner Affairs
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
, the NRC may not conduct or sponsor, and a Qlt***""*"'
person is not required to respond to, the information collection.
- 3. Page Peach Bottom Atomic Power Station Unit 3 05000278 1
OF 4
- 4. Title Reactor Coolant System Pressure Boundary Leakage Resulting in Technical Specification Required Shutdown
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved I
Sequential I Rev Facility Name Docket Number Month Day Year Year
,, Number No.
Month Day Year 05000 Facility Name Docket Number 09 21 2018 2018 -
002 1
12 12
- 2018 05000
- 9. Operating Mode
- 11. This Report is Submitted Pursuant to the Requirements of 1 O CFR §: (Check all that apply)
D 20.2201(b)
- - D 20.2203(a)(3)(i)
IZJ, 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201(dl D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B) 1 D 20.2203(a)(1)
D D
D 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)
D 20.22os(a)(2)(il D 50.36(c)(1 )(i)(A) o* 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x) 1 O. Power Level D 20.22os(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(5)
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1) 100%
D 20.2203(a)(2)(v)
- IZJ 50.73(a)(2)(i)(A)
IZI 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
IZI 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
';' *.t.jiti.>.c D
D
,.A"'
50.73(a)(2)(i)(C)
Other (Specify in Abstract below or in =
Unit Conditions Prior to the Event YEAR 2018 SEQUENTIAL NUMBER 002 Unit 3 was operating in Mode 1 at approximately 100% rated thermal power. There were no structures, systems or components out of service that contributed to this event.
Description of Event
REV NO.
1 On 9/20/18 at approximately 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />, an Equipment Operator, during normal rounds, identified that differential pressure switches DPIS-3-23-076 and DPIS-3-23-077 [EIIS:PDS] for the High Pressure Coolant lrijectio_n (HPCI) system [EIIS:BJ] were reading lower than expected.
Upon further review and investigation, it was determined that the switches would-not have been able to perform their Technical Specification (TS) 3.3.6.1 required function of isolating the HPCI steani line in the event of a break. The two switches were declared inoperable on 9/21 /18 at approximately 1755 hours0.0203 days <br />0.488 hours <br />0.0029 weeks <br />6.677775e-4 months <br /> and were inoperable for approximately 22-1/2 hours as a result of this event. Required actions of TS 3.3.6.1 were performed, including isolation of HPC.1 steam line. As a result of the isolation, TS 3.5.1 Condition C was entered for HPCI being inoperable, which requires the system to be restored within 14 days.
Further analysis of the low differential pressure indicated on the DPISs suggested a possible leak of the instrument line upstream of the DPISs. This was supported by a small increase in measured drywell leakage that was identified on 9/20/18. A reactor power reduction to 35% was initiated on 9/22/18 at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> to support an entry into the drywell. The drywell inspection verified leakage of the 1-inch diamet.er instrument line (Schedule 80, Type 304 stainless steel), which is a reactor coolant system pressure boundary. TS 3.4.4 Condition C was entered at 0955 hours0.0111 days <br />0.265 hours <br />0.00158 weeks <br />3.633775e-4 months <br /> on 9/22/18, which requires the unit to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A controlled shutdown was initiated and the unit entered Mode 3 at 2029 hours0.0235 days <br />0.564 hours <br />0.00335 weeks <br />7.720345e-4 months <br /> on 9/22/18 and Mode 4 at 0701 hours0.00811 days <br />0.195 hours <br />0.00116 weeks <br />2.667305e-4 months <br /> on 9/23/18.
The lea!< was located at a point where the instrument line was supported by a rectangular shaped bracket.
Vibration had caused the bracket to wear the instrument line, eventually resulting in a through-wall steam leak. The section of pipe that contained the leak was replaced. Similar small-bore piping was inspected and repairs performed as necessary.
Following repairs, a normal startup was commenced and Unit 3 entered Mode 1 at 1638 hours0.019 days <br />0.455 hours <br />0.00271 weeks <br />6.23259e-4 months <br /> on 9/25/18.
Analysis of Event
This event is being reported in accordance with the following:
1 O CFR 50.73(a)(2)(i)(A) - the completion of any nuclear plant shutdown required by the plant's Technical Specifications. TS 3.4.4 Condition C;reqb.lires the unit to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for any reactor coolant-system pressure boundary leakage. This TS action was entered on 9/22/18 at 0955 hours0.0111 days <br />0.265 hours <br />0.00158 weeks <br />3.633775e-4 months <br /> when leakage was identified on an instrument line associated with monitoring HPCI steam line differential pressure. The instrument line is part of the RCS pressure boundary.
1 O CFR 50.73(a)(2)(i)(B) - any condition prohibited by plant Technical Specifications. Two conditions prohibited by TS existed:
"1
- 1) Differential pressure switches DPIS-3-23-076 and DPIS-3-23-077 were identified as reading low on 9/20/18 at approximately 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />. Further analysis determined that they would not have been able to perform their safety function of isolating the HPCI turbine steam line in the event of a steam line break. TS 3.3.6.1 Condition B requires the isolation capability performed by these switches to be restored in one hour. If not restored. within one hour, Condition F requires the penetration flow path to be isolated within one hour (two hours after discovery). Because of the additional analysis performed to determine if they could perform their safety function, the required actions were not performed until 9/21/18 at approximately 1755 hours0.0203 days <br />0.488 hours <br />0.0029 weeks <br />6.677775e-4 months <br />, which is approximately 22-1/2 hours after discovery.
- 2) Differential pressure switches DPIS-3-23-076 and DPIS-3-23-077 were identified as reading low on 9/20/18 at approximately 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />. This was not known to be due to pressure boundary leakage until 9/22/18 at 0955 hours0.0111 days <br />0.265 hours <br />0.00158 weeks <br />3.633775e-4 months <br />, at which time TS 3.4.4 Condition C was entered. TS 3.4.4 Condition C requires the unit to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for any pressure boundary leakage. The unit entered Mode 3 at 2029 hours0.0235 days <br />0.564 hours <br />0.00335 weeks <br />7.720345e-4 months <br /> on 9/22/18 and Mode 4 at 0701 hours0.00811 days <br />0.195 hours <br />0.00116 weeks <br />2.667305e-4 months <br /> on 9/23/18. Because there is firm evidence that pressure boundary leakage existed prior to the time of discovery on 9/22/18, the TS required completion times were not met. Mode 3 was entered approximately 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> and Mode 4 was entered approximately 59 hours6.828704e-4 days <br />0.0164 hours <br />9.755291e-5 weeks <br />2.24495e-5 months <br /> from the time low readings were identified on the DPISs.
1 O CFR 50.73(a)(2)(ii)(A) - the condition of the plant, including its principal safety barriers, being seriously degraded. Leakage was identified on an instrument line associated with monitoring HPCI steam line differential pressure. The instrument line is part of the RCS pressure boundary, which is
- one of the principal safety barriers.
1 O CFR 50.73(a)(2)(v)(D) - any condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident. HPCI is considered a single train system and was required by TS to be declared inoperable as a result of this event.
Cause of the Event
The one-inch diameter instrument line was aligned such that it was in contact with one side of a rectangular support. Vibration-induced wear (fretting) reduced the wall thickness of the pipe at the contact point until a steam leak developed. It could not be determined why the pipe was aligned such that it was in contact with the support bracket.
Corrective Actions
The section of pipe where the leak occurred was replaced. Inspections were performed on other small-bore piping in the drywell that are supported by the same st¥1~upport bracket and subject to vibration. One other location was identified where fretting was significant enough to proactively have the pipe replaced.
SEQUENTIAL NUMBER 002 REV NO.
1 Piping with less significant fretting was identified at nine other locations. These wear areas were wrapped with a layer of stainless steel sheet metal to protect the line until permanent repairs can be completed if needed. Additional corrective actions are documented in the corrective action program.
Previous Similar Occurrences No previous similar occurrences have been identified of RCS pressure boundary leakage due to fretting.
Unit 3 LER 17-001 documents RCS pressure boundary leakage due to a weld failure in a one-inch diameter instrument line. Unit 3 LE~ 05-003 documents RCS pressure boundary leakage due to a weld failure in a one-inch diameter equalizing line in the Residual Heat Removal System. Page _4_ of...!...