05000338/LER-2014-003, Regarding Technical Specification Required Shutdown Due to Reactor Coolant System Pressure Boundary Leakage
| ML15058A035 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/11/2015 |
| From: | Gerald Bichof Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 14-619 LER 14-003-00 | |
| Download: ML15058A035 (5) | |
| 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 |
| 3382014003R00 - NRC Website | |
text
1 OCFR50.73 Virginia Electric and Power Company North Anna Power Station 1022 Haley Drive Mineral, Virginia 23117 February 11, 2015 Attention: Document Control Desk Serial No.:
14-619 U. S. Nuclear Regulatory Commission NAPS:
RAP Washington, DC 20555-0001 Docket No.: 50-338 License No.: NPF-4
Dear Sirs:
Pursuant to 10CFR50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to North Anna Power Station Unit 1.
Report No. 50-338/2014-003-00 This report has been reviewed by the Facility Safety Review Committee and will be forwarded to the Management Safety Review Committee for its review.
Sincerely, Gerald T. Bischof Site Vice President North Anna Power Station Enclosure Commitments contained in this letter: None cc:
United States Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station
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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017
'02-2014)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections LI ENSE EVENT RE T LBranch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by L C N E E
REPORT~i i(Jr'.Jii IE
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internet e-mail to Infocollects.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 20503. If a means used to impose an information collection does not display a currently valid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE North Anna Power Station 05000338 1 OF 4
- 4. TITLE Technical Specification Required Shutdown due to Reactor Coolant System Pressure Boundary Leakage
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FCLIYNMEDCETNME NUMBER NO.
FACILITY NAMvE DOCKET NUMIBER 12 22 2014 2014 -
003 00 02 11 2015 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
E] 20.2201((b)
E] 20.2203(a)(3)(i)
[]
50.73(a)(2)(i)(C)
[]
50.73(a)(2)(vii)
EI 20.2201(d)
El 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
E] 20.2203(a)(4)
F] 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[]
20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
E] 20.2203(a)(2)(iii)
Li 50.36(c)(2)
LI 50.73(a)(2)(v)(A)
El 73.71(a)(4)
E] 20.2203(a)(2)(iv)
E] 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.7 1(a)(5) 30 El 20.2203(a)(2)(v)
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50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER t...l-"l20.203()(2)vi)r-I
'm]Specify in Abstract below or in l202203(a)(2)(v)
E 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
NRC beu 3r6A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT TELEPHONE NUMBER (Include.Area Code)
Gerald T. Bischof, Site Vice President
- .(540) 894-2101CAUSE SYSTEM COMPONENT MANU-REPORTABLE Il MANU-REPORTABLE FACTURER TO EPIX
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION 0] YES (If yes. complete 15. EXPECTED SUBMISSION DATE)
[]
NO DATEF ABSTRACT (Limit to 1400 spaces, i.e., approximatelv 15 single-spaced typewritten lines)
On December 22, 2014, Unit 1 reactor power was reduced from 100% to 30% to allow entry into the Reactor Coolant System loop rooms to investigate an increased unidentified leak rate of 0.053 gallons per minute. At 2230 on December 22, 2014, with Unit 1 operating at 30% power, during a containment walkdown, steam was discovered coming from underneath the lagging on the "B" Reactor Coolant System intermediate loop. Further investigation identified a pressure boundary leak on the "B" loop drain piping between the loop connection and 1-RC-68, the "B" Loop Cold Leg Drain Isolation Valve. At that time, the limiting action of Technical Specification 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 />. Due to the pressure boundary leakage, this event was reported at 0057 on December 23, 2014, in accordance with 10 CFR 50.72(b)(2), 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." The health and safety of the public were not affected by this event.
NRC FORM 366 (02-2014)
1.0 DESCRIPTION OF THE EVENT On December 22, 2014, Unit I reactor power was reduced from 100% to 30% to allow entry into the Reactor Coolant System (RCS) (EIIS System - AB) loop rooms to investigate an increased unidentified leak rate of 0.053 gallons per minute (gpm). At 2230 on December 22, 2014, with Unit 1 operating at 30% power, during a containment walkdown, steam was discovered coming from underneath the lagging on the "B" RCS intermediate loop. Further investigation identified a pressure boundary leak was on the "B" loop drain piping between the loop connection and 1-RC-68, the "B" Loop Cold Leg Drain Isolation Valve (EIIS System - AB, Component - V). 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 />. At 2345 on December 22, 2014, the power ramp to remove the unit from service commenced. At 0226 on December 23, 2014, the unit was placed in Mode 3 and the unit was placed in Mode 5 at 1629 on December 23, 2014.
Due to the pressure boundary leakage, this event was reported at 0057 on December 23, 2014, in accordance with 10 CFR 50.72(b)(2), 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."
After Unit 1 was in Mode 5, further investigation, via Ultrasonic Testing (UT), of the other loop drain lines found that there was an indication on the corresponding "C" cold leg loop drain elbow, but no leakage was noted. Subsequently, this indication was evaluated as acceptable, per ASME code, until the next refueling outage in the spring of 2015.
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
No significant safety consequences resulted from this event because Unit 1 was promptly removed from service and the "B" cold leg loop drain elbow was replaced. The health and safety of the public were not affected by this event.
3.0 CAUSE
The direct cause of the through-wall leak in the "B" loop drain line elbow was a thermal fatigue failure of the elbow. Limitations in the industry generic model used to predict swirl penetration thermal fatigue in stagnant RCS branch lines allowed the Non-
Destructive Examination (NDE) test frequency of the "B" Loop drain line elbow to be set non-conservatively, causing thermal fatigue cracking to go unmonitored. The Root Cause is a legacy issue from 2009 due to the generic model, as developed by EPRI, not accurately predicting fatigue damage due to swirl penetration.
A contributing cause was the fact that Unit 1 chemistry samples of the RCS have been collected from the "B" cold leg loop drain line since 2004. Chemistry sampling added additional thermal stresses which accelerated the fatigue failure in the "B" loop drain line elbow. The screening calculation performed in 2010 to identify lines susceptible to thermal fatigue was performed by Site and Corporate Engineering. The screeners did not have a strong questioning attitude to identify chemistry sampling as a thermal fatigue contributor. A questioning attitude would have prompted a discussion with Chemistry personnel for additional information on sample duration and frequency. In 2011, the review process for calculations and other engineering products was improved after Engineering issued CM-AA-ECR-101, Engineering Challenge Reviews (ECR). This Guidance and Reference Document (GARD) improves the accuracy and completeness of engineering products by having a diverse review team. Engineering calculations are listed for consideration for and ECR at the discretion of Engineering Management.
4.0 IMMEDIATE CORRECTIVE ACTION(S)
The drain line elbow was replaced while the unit was in Mode 5.
5.0 ADDITIONAL CORRECTIVE ACTIONS
The Unit 1 "C" cold leg loop drain elbow will be replaced during the Unit 1 spring 2015 refueling outage. Engineering will then perform a Material Analysis of the elbows and implement any necessary changes to the Thermal Fatigue Management Program.
6.0 ACTIONS TO PREVENT RECURRENCE Baseline UT will be performed on all Unit 1 and Unit 2 hot and cold leg drain line elbows. The North Anna Augmented Inspection Program, ER-NA-AUG-101, will be revised to perform a UT examination on the cold leg drain elbows based upon guidance found in EPRI MRP-146S for locations where significant thermal fatigue is predicted.
Additionally, Engineering will work with EPRI and industry peers to develop a new model and/or new guidance to better predict the impact of thermal fatigue and other sources of thermal loading.
7.0 SIMILAR EVENTS
No similar events have occurred at North Anna.
8.0 ADDITIONAL INFORMATION
Unit 2 continued operating in Mode 1,100 percent power during this event.