05000339/LER-2016-001, Regarding Technical Specification Required Shutdown Due to Reactor Coolant System Leak
| ML16271A408 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/21/2016 |
| From: | Gerald Bichof Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 16-001-00 | |
| Download: ML16271A408 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) |
| 3392016001R00 - NRC Website | |
text
10CFR50.73 Virginia Electric and Power Company North Ann~ Power Station 1022 Haley Drive Mineral, Virginia 23117 September 21, 2016 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Sirs:
Serial No.:
16-316 NAPS:
DPM Docket Nos.: 50-339 License Nos.: NPF-7 Pursuant to 10CFR50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to North Anna Power Station Unit 2.
Report No. 50-339/2016-001-00 This report has been reviewed by the Facility Safety Review Committee and will be forwarded to the Management Safety Review Committee for its review.
Enclosure Commitments contained in this letter: None Sincerely, 1-5~?&-."6 Gerald T. Bischof Site Vice President North Anna Power Station cc:
United States Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 NRG Senior Resident Inspector North Anna Power Station
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSlm APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 (11-2015)
Estimated burden per response to comply -Mth this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
c:J>'Uf4>4
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Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and lnfonmation Collections LICENSEE EVENT REPORT (LER)
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC digits/characters for each block) 20503. If a means used to impose an infonmation collection does not display a currently valid OMB control number, the NAG may not conduct or sponsor, and a person is not required to respond to, the infonmation collection.
- 3. PAGE North Anna Power Station, Unit 2 05000339 1 OF 4
- 4. TITLE Technical Specification Required Shutdown Due to Reactor Coolant System Leak
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 07 30 2016 2016 - 001 -
00 09 21 2016 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.2201(b)
D 20.2203(a)(3J(il
~ 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) 1 D
D 20.2203(a)(4)
D D
20.2203(a)(1)
- 50. 73(a)(2)(iii)
- 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)
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
~
D D
20.2203(a)(2)(v)
- 50. 73(a)(2)(i)(A)
- 50. 73(a)(2)(v)(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)(2J(ii)
D 50.73(a)(2)(i)(C)
D OTHER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONT ACT I
TELEPHONE NUMBER (Include Area Code}
Gerald T. Bischof, Site Vice President (540) 894-2101 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B
AB PSX y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On July 29, 2016, with Unit 1 and Unit 2 at 100 percent power in mode 1, Unit 2 Reactor Coolant System (RCS) unidentified leakage was noted to take a 0.05 gallons per minute step increase coincident with a similar increase in Unit 2 containment sump in-leakage. Subsequent containment entries identified a through wall leak existed in the controlled bleed-off piping associated with the Reactor Coolant Pump seal for 2-RC-P-1 C. Due to the pressure boundary leakage, this event was reported at 1517 hours0.0176 days <br />0.421 hours <br />0.00251 weeks <br />5.772185e-4 months <br /> on July 30, 2016, in accordance with 10 CFR 50.72(b)(2)(i), for "The initiation of any nuclear plant shutdown required by the plant's Technical Specifications" and 10 CFR 50.72(b)(3)(ii)(A) for "Any event or condition that results in the condition of the nuclear power plant including its principle safety barriers, being seriously degraded." While in Mode 5, the controlled bleed-off piping associated with the RCP seal for 2-RC-P-1 C was replaced. The health and safety of the public were not affected by this event.
NRC FORM 366 (11-2015)
Page 2 of 4 1.0 DESCRIPTION OF THE EVENT On July 29, 2016 Reactor Coolant System (RCS) (EllS System - AB) unidentified leakage was identified to take a 0.05 gallons per minute step increase coincident with similar increase in Unit 2 containment sump in-leakage.
Containment entries were initiated to determine the source of the leakage. On July 30, 2016, at 1152 hours0.0133 days <br />0.32 hours <br />0.0019 weeks <br />4.38336e-4 months <br />, it was determined that there was an un-isolatable through wall leak in the controlled bleed-off piping (EllS System - AB, Component - PSX) associated with the Reactor Coolant Pump (RCP) (EllS System - AB, Component - P) seal for 2-RC-P-1 C. At that time, the limiting action of Technical. Specification (TS) 3.4.13, RCS Operational Leakage, Condition B was entered which required placing the unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, and the power ramp to remove the Unit 2 from service commenced. At 1702 on July 30, 2016, the unit was placed in Mode 3 and the unit was placed in Mode 5 at 0538 on July 31, 2016.
Due to the pressure boundary leakage, this event was reported* at 1517 hours0.0176 days <br />0.421 hours <br />0.00251 weeks <br />5.772185e-4 months <br /> on July 30, 2016, in accordance_ with 10 CFR 50.72(b)(2)(i), for "Initiation of plant shutdown required by Technical Specifications" and 10 CFR 50.72(b)(3)(ii)(A) for "Any event or condition that results in the condition of the nuclear power plant including its principle safety barriers, being seriously degraded."
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
No significant safety consequences resulted from this event because Unit 2 was promptly removed from service and the affectecj controlled bleed.:.off piping associated with the RCP seal for 2-RC-P-1 C was replaced. The health and safety of the public were n*ot affected by this event.
3.0 CAUSE
The. direct cause of the RCS unidentified leakage was determined to be a large mean stress.placed on the socket weld due to the controlled bleed-off line not being properly aligned in the. downstream pipe support, and therefore not allowing for the thermal growth of the RCS. As a result of the large mean stress, a crack initiated at a small
- defect (lack of fusion) in the toe of the socket weld and propagated through the weld due to normal cyclic vibration-from the. Reactor Coolant Pump.
- NRC FORM 366A (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET North Anna Power Station, Unit 2 05000339 YEAR
'6. LER NUMBER SEQUENTIAL NUMBER REV NO.
2016 -
001 00 The root cause was determined to be barriers in the implementation of a Design Change (DC), for replacement of the Unit 2 Reactor Coolant Pump seals, which were insufficient for personnel to consider and include pipe supports outside the modification boundary
- into th.e work scope and subsequent inspections. E:ngineering and the implementing craft missed opportunities to identify and inspect for proper pipe alignment in the downstream pipe support.
A contributing cause was the fact that design decisions made in determining scope for the DC were not adequately validated prior to the design change being issued and implemented, resulting in the pipe support being considered outside the modification boundary. Changes to the piping configuration were assumed to have no effect on the downstream piping; thus, no safeguards were put in place to ensure that the pipe would not move within the downstream pipe support.
As* a result, potential unintended movement of the pipe in the support was not identified during inspections supporting the design change. During implementation, the craft decided that the piping downstream of the work scope would not be adversely impacted by fit-up of the modified seal return line. No pre,..job or post-job walkdown was performed to validate this assumption.
4.0 IMMEDIATE CORRECTIVE ACTION(S)
While in Mode 5, the controlled bleed-off piping associated with the RCP seal for 2-RC-P-1 C was replaced. Other welds on the seal injection and seal return line~ of all three Unit 2 RCPs were inspected with no signs of cracks or defects. A natural frequency (or_
bump test) and/or vibration data collection was performed on all three seal return lines to determine if a resonance condition exists with no issues found.
5.0 ADDITIONAL CORRECTIVE ACTIONS
Design Engineering will review a sample set of design changes for Unit 1 and Unit 2 that involve modifications fo piping, piping components, and pipe supports, particularly those near *rotating equipment. During the appropriate Refueling Outage, a walkdown will be conducted of the sampled modifications to verify correct pipe and pipe support*
configuration, paying particular attention to supports outside of the modification boundary.
Page 4 of 4 (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME 2."DOCKET NUMBER North Anna Power Station, Unit 2 05000339 YEAR
- 6. LER NUMBER SEQUENTIAL NUMBER REV NO.
2016 -
001 00 6.0 ACTIONS TO PREVENT RECURRENCE Construction and maintenance implementing procedures will be revised to ensure supports upstream and downstream of the work scope are addressed for any work performed on piping, piping components or supports, unless otherwise approved by engineering.
During the Fall 2016 Unit 1 refueling outage, extent of condition inspections, measurements, and vibration readings will be taken on the Unit 1 RCP seal lines and supports.
A walkdown of the component cooling. lines to the RCP thermal barriers to verify alignment.of piping supports will be conducted during the Fall 2016 refueling outage for Unit 1 and the Fall 2017 refueling outage for Unit 2.
7.0 SIMILAR EVENlS No similar events have occurred at North Anna.
8.0 MANUFACTURER/MODEL NUMBER N/A
9.0 ADDITIONAL INFORMATION
Unit 1 continued operating at 100 percent power, mode 1, during this event.