05000260/LER-2010-001, For Browns Ferry, Unit 2 Regarding Condition Prohibited by Technical Specifications When Two Emergency Core Cooling Systems, Loops I and II of the Residual Heat Removal System Low Pressure Coolant Injection System, Become Inope

For Browns Ferry, Unit 2 Regarding Condition Prohibited by Technical Specifications When Two Emergency Core Cooling Systems, Loops I and II of the Residual Heat Removal System Low Pressure Coolant Injection System, Become Inoperable

ML101180251
Person / Time
Site:	Browns Ferry
Issue date:	04/26/2010
From:	Polson K Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 10-001-00
Download: ML101180251 (8)

LER-2010-001, For Browns Ferry, Unit 2 Regarding Condition Prohibited by Technical Specifications When Two Emergency Core Cooling Systems, Loops I and II of the Residual Heat Removal System Low Pressure Coolant Injection System, Become Inoperable

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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
2602010001R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 April 26, 2010 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Browns Ferry Nuclear Plant, Unit 2 Facility Operating License No. DPR-52 NRC Docket No. 50-260

Subject:

Licensee Event Report 50-26012010-001-00 The enclosed Licensee Event Report provides details of a condition prohibited by technical specifications when two emergency core cooling systems, Loops I and II of the Residual Heat Removal System Low Pressure Coolant Injection System, became inoperable. The Tennessee Valley Authority is submitting this report in accordance with 10 CFR 50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant's Technical Specifications.

There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact S. T. Day, Acting Site Licensing and Industry Affairs Manager, at (256) 729-2636.

Respectfully, J.Poison U Vice President cc: See page 2

U.S. Nuclear Regulatory Commission Page 3 April 26, 2010 STD:MWO:LAJ Enclosure bcc (Enclosure):

G. P. Arent, EQB 1B-WBN T. E. Cribbe, LP 4K-C D. E. Jernigan, LP 3R-C L. A. Jones, SAB 2B-BFN R. M. Krich, LP 3R-C J. H. McCarthy, NAB 1A-BFN K. J. Poison, NAB 2A-BFN J. J. Randich, POB 2C-BFN P. D. Swafford, LP 3R-C L. E. Thibault, LP 3R-C E. J. Vigluicci, WT 6A-K INPO: LEREvents@inpo.org NSRB Support, LP 5M-C EDMS, WT CA-K File S:\\Licensing\\LIC\\EVERYONE\\2010\\Submittals\\LER\\LER 50-260 2010-001 Rev 0.pdf

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 08/31/2010 (9-2007)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information (See reverse for required number of collection.

3. PAGE Browns Ferry Nuclear Plant Unit 2 05000260 1 of 6

4. TITLE: Condition Prohibited By Technical Specifications When Two Emergency Core Cooling Systems, Loops I and II of the Residual Heat Removal System Low Pressure Coolant Injection System, Became Inoperable

6. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MO YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR N/A N/A NUMBER NO.

FACILITY NAME DOCKET NUMBER 02 25 2010 2010 -

001 00 04 26 2010 N/A N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b)

[

20.2203(a)(3)(i)

[I 50.73(a)(2)(i)(C)

[I 50.73(a)(2)(vii) 10 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A)

[] 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 0] 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii)

El 50.73(a)(2)(ix)(A)

10. POWER LEVEL 0 20.2203(a)(2)(ii)

E0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x) o 20.2203(a)(2)(iii) 0 50.36(c)(2) 0l 50.73(a)(2)(v)(A) 0 73.71 (a)(4)

El 20.2203(a)(2)(iv) 0l 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) 0 73.71 (a)(5) 100 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A)

[I 50.73(a)(2)(v)(C) 0 OTHER El 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0l 50.73(a)(2)(v)(D)

Speci isnAbstat be.r in3NRC Ferm, 356A

12. LICENSEE CONTACT FOR THIS LER NAME TELEPHONE NUMBER (Include Area Code)

Mike Oliver, Licensing Engineer 256-729-7874CAUSE S

MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (if yes, complete 15. EXPECTED SUBMISSION DATE)

NO DATE N/A N/A N/A ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced ypewritten lines)

Back leakage through Emergency Core Cooling System (ECCS) Residual Heat Removal (RHR) Loop II low pressure coolant injection (LPCI) valves resulted in the water in the pipe upstream of the inboard valve to be at an elevated temperature with respect to ambient conditions. To prevent voiding in the piping, monitoring with trigger points/actions were established. RHR Loop I was inoperable for scheduled maintenance. On February 25, 2010, at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br /> Central Standard Time, a trigger point was met, and Operations personnel entered Technical Specifications (TS) Limiting Condition for Operation (LCO),3.0.3 based on entry into TS LCO 3.5.1 Condition H when RHR Loops I and II were inoperable.

Operations personnel began lowering reactor power at 1935 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.362675e-4 months <br /> because the conditions requiring entry into TS LCO 3.0.3 could not be corrected within an hour. TS LCOs were exited at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> when RHR Loop II system pressure was restored to ensure that the piping would remain filled. At 2320 hours0.0269 days <br />0.644 hours <br />0.00384 weeks <br />8.8276e-4 months <br />, reactor power was returned to 100 percent.

The cause of this event is aged components and lack of preventive maintenance activities on valve internals.

The Tennessee Valley Authority is submitting this report in accordance with 10 CFR 50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant's Technical Specifications.

NRC FORM 366 (9-2007)

(If more space is required, use additonal copies of (If more space is required, use additional copies of (If more space is required, use addidonal copies of (If more space is required, use additional copies of NRC Form 366A) (17)

V. ASSESSMENT OF SAFETY CONSEQUENCES

Monthly venting surveillances are performed on all Loops of RHR, and there have been no indications of increased temperature (by observance of steam) on the Unit 2 Loop I or on another BFN unit. However, it was determined that the extent of condition includes the LPCI outboard and inboard valves on RHR Loops I and II on all three Units.

The safety consequences of this event were not significant. It was determined that this condition does not compromise the primary containment isolation function or any operational mode of Unit 2 RHR Loop II. TVA concludes that the Unit 2 RHR system is capable of performing its design function for as long as the temperature in the piping is maintained below that prescribed in the ODMI 210437. Further, ISI UT before and after the TS LCO entry confirmed the absence of voids. The condition was also evaluated as being neither degraded nor non-conforming.

Therefore, TVA concludes that there was no significant reduction in the protection of the public by this event.

VI. CORRECTIVE ACTIONS

A.

Immediate Corrective Actions

In accordance with ODMI guidance and Operating Instruction 2-01-74, RHR System, Operations personnel aligned the Unit 2 ECCS keep fill from the PSC head tank to the CS&S System to raise the system pressure to ensure that no steam voiding occurred in the Loop II LPCI piping. This action placed Loop II in an acceptable condition with respect to the applicable ODMI 210437 trigger point and action.

B.

Corrective Actions to Prevent Recurrence Actions from the causal analysis are, for all three BFN units, 1) to revise the RHR System Monitoring Plan to incorporate taking periodic pipe temperatures of LPCI piping and 2) to initiate PMs to inspect/refurbish the intemals of the LPCI check and inboard isolation valves.

VII. ADDITIONAL INFORMATION

A.

Failed Components None B.

PREVIOUS LERS ON SIMILAR EVENTS None C.

Additional Information

Corrective action documents for this report are PERs 218493 and 210437.

D.

Safety System Functional Failure Consideration:

This event is not classified as a safety system functional failure according to NEI 99-02. In view of the fact that RHR Loop II remained functional, this event is not reportable under 10 CFR 50.73(a)(2)(ii)(B), (unanalyzed condition), or 10 CFR 50.73(a)(2)(v)(B) (removal of residual heat) and (D) (mitigate the consequences of an accident).U.S. NUCLEAR REGULATORY COMMISSION (9-2007)

LICENSEE EVENT REPORT (LER)

FACILITY NAME (1)

DOCKET (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Browns Ferry Nuclear Plant Unit 2 05000260 2010

-- 001

-- 00 6 of 6 NAKXRIA iVE (It more space is required, use additional copies of NRC Form 366A) (17)

E.

Scram With Complications Consideration:

This event did not include a reactor scram.

VIII. COMMITMENTS

None