05000266/LER-1917-003, Regarding Degraded Condition
| ML17347A773 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/13/2017 |
| From: | Coffey B Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2017-0054 LER 17-003-00 | |
| Download: ML17347A773 (3) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(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) |
| 2661917003R00 - NRC Website | |
text
December 13, 2017 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Unit 1 Docket 50-266 Renewed License Nos. DPR-24 Licensee Event Report 266/2017-003-00 NRC 2017-0054 10 CFR 50.73 Enclosed is Licensee Event Report (LER) 266/2017-003-00 for Point Beach Nuclear Plant, Unit 1. NextEra Energy Point Beach, LLC is providing this LER regarding the degraded condition on Unit 1.
This letter contains no new regulatory commitments.
If you have any questions please contact Mr. Eric Schultz, Licensing Manager, at 920/755-7854.
Sincerely, NextEra Energy Point Beach, LLC Bob Coffey Site Vice President Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, Wl54241
NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017) httg://www. nrc.gov/read ing-rm/doc-collections/nu regs/staff/sr1 022/r3D the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
.PAGE Point Beach Nuclear Plant Unit 1 05000266 1 of 2
- 4. TITLE Degraded Condition
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR NA NA 10 30 2017 2017 003 00 12 13 2017 FACILITY NAME DOCKET NUMBER NA NA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: Check all that apply)
. MODE 3 D 20.2201(b)
D 20.2203(a)(3)(i)
~ 50. 73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201(d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(iii)
D 50. 73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii) 0 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) 0%
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii) 0 50.73(a)(2)(i)(C) 0 OTHER Specify in Abstract below or In NRC Fonn 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT
~~LEPHONE NUMBER (Include Area Code)
Thomas P. Schneider, Senior Licensing Engineer 920-755-7797 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TO EPIX B
AB PSF NA y
NA NA NA NA NA
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
[ZI NO SUBMISSION NA NA NA DATE
~BSTRACT (Limit to 1400 spaces, Le., approximately 15 single-spaced typewritten lines)
On October 30, 2017, with Unit 1 in MODE 3 for refueling activities, a boric acid indication downstream of 1CV-309B, 1 P-1 8 Reactor Coolant Pump (RCP) Labyrinth Seal 1 DPT-124 Upper Root Valve was identified as a through-wall flaw. The flaw location on the root valve to differential pressure transmitter (OPT) instrument tubing welded joint was within the reactor coolant system (RCS) pressure boundary.
This event is being reported pursuant to 10 CFR 50. 73(a)(2)(ii)(A) for material defects in the primary coolant system that were not acceptable in accordance with ASME Section XI.
NRC FORM 366 (04-2017)
Description of the Event:
SEQUENTIAL NUMBER 003 REV NO.
00 At 0404 CST on October 30, 2017, with Unit 1 in MODE 3 for refueling activities, a boric acid indication downstream of 1 CV-3098, 1 P-1 B RCP Labyrinth Seal 1 DPT-124 Upper Root Valve was identified as a through-wall flaw. The flaw location on the root valve [RTV] to DPT [PDT] instrument tubing welded joint [PSF] was within the RCS [AB]
pressure boundary.
The root valve to OPT instrument tubing welded joint has been repaired and returned to service.
This 60 day licensee event report is being submitted in accordance with the requirements of 10 CFR 50. 73(a)(2)(ii)(A) for any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being degraded. The material defect in the primary coolant system could not be found acceptable in accordance with ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws" or ASME Section XI, Table IWB-3410-1, "Acceptance Standards."
Cause of the Event
The most likely cause of the degraded barrier was crater cracking at a weld stop point in the root valve to instrument tubing welded joint.
Analysis of the Event
Maintenance activities identified a through-wall flaw downstream of the root valve to the RCP Labyrinth Seal OPT instrument tubing welded joint. The condition was discovered during reactor startup activities after the reactor had been shut down for several weeks for a planned refueling. The labyrinth seal serves as a buffering interface, to limit the exchange of reactor coolant from the seal portion of the RCP. The RCP Labyrinth Seal OPT instrument provides performance information only of the buffer interface. The through-wall flaw was within the RCS pressure boundary.
The through-wall flaw in the welded joint was most likely the result of crater cracking at a weld stop point. The flaw has been corrected. Adequate heat removal capability for the reactor was provided by two separate engineered safety feature system trains.
Corrective Actions
The root valve to OPT instrument tubing welded joint has been repaired and returned to service. RCS pressure boundary systems and components will continue to receive repetitive visual examinations on predetermined frequencies in accordance with existing programs.
Safety Significance
The event was determined to be of very low safety significance. The flaw location was on small bore piping. There was no loss of any safety systems, structures or components needed to shut down the reactor, maintain safe shutdown conditions, remove residual heat, control the release of radioactive material or mitigate the consequences of an accident. There was no impact on the health and safety of the public as a result of this condition.
Similar Events
There have not been similar events of this degraded condition in the past three years.
Component Failure Data
None Page 2 of 2