05000334/LER-2010-002-01, Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld
| ML110750024 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley (DPR-066) |
| Issue date: | 03/11/2011 |
| From: | Harden P FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-11-061 LER 10-002-01 | |
| Download: ML110750024 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3342010002R01 - NRC Website | |
text
FENOC Beaver Valley Power Station P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Paul A. Harden 724-682-5234 Site Vice President Fax: 724-643-8069 March 11, 2011 L-1 1-061 10 CFR 50.73 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 LER 2010-002-01 Enclosed is Licensee Event Report (LER) 2010-002-01, "270 Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld." This event was previously reported in accordance with 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B) on November 29, 2010. This LER Supplement updates the direct cause of event based on the results of the metallurgical examination of residual heat removal system~drain valve socket weld failure.
There are no regulatory corfnmitments contained in this submittal. Any actions discussed in this document that represenht.intended or planned actions are described for the NRC's information, and are not regulatory commitrmients.
If there are any questions or if additional information is required, please *contact Mr. Brian T. Tuite, Manager, Regulatory Compliance at 724-682-4284.
- Sincerel, Paul A. Harden Attachment c:
Mr. W. M. Dean, NRC Region I Administrator Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager INPO Records Center (via electronic image)
Mr. L. E. Ryan (BRP/DEP)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/31/2013 (10-2010)
, the NRC may digits/characters fnot conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Beaver Valley Power Station Unit Number 1 05000334 1 of 5
- 4. TITLE 270 Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YER YEAR NUMBER NO.
MONTH DAY YEAR None FACILITY NAME DOCKET NUMBER 10 02 2010 2010 002 01 03 11.2011
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 5 20.2201(b)
E] 20.2203(a)(3)(i)
M 50.73(a)(2)(i)(C)
E] 50.73(a)(2)(vii) 5
[-] 20.2201 (d)
E] 20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
E] 20.2203(a)(4)
E] 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[_] 20.2203(a)(2)(i)
E] 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii) 0l 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[] 20.2203(a)(2)(ii)
[] 50.36(c)(1)(ii)(A)
[] 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
E] 20.2203(a)(2)(iii)
E] 50.36(c)(2)
E] 50.73(a)(2)(v)(A)
[E 73.71(a)(4) 0 %
[
20.2203(a)(2)(iv)
E] 50.46(a)(3)(ii)
[
50.73(a)(2)(v)(B)
LI 73.71(a)(5)
[
20.2203(a)(2)(v)
E] 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
[] OTHER D
Specify in Abstract below E] 20.2203(a)(2)(vi)
Z 50.73(a)(2)(i)(B)
E] 50.73(a)(2)(v)(D) or in Following offload of the reactor core, the RHR System was no longer required to be in operation and was isolated to repair 1 RH-200. 1 RH-200 was permanently repaired by replacement on October 15, 2010: 1RH-200 was declared operable on October 18, 2010 following completion of required post-maintenance inspections/testing.
CAUSE OF EVENT
The direct cause is fatigue failure of the valve inlet socket weld on 1 RH-200. The most probable root cause is less than adequate design considerations to accommodate vibration induced fatigue on small bore vents and drains in the BVPS Unit No. 1 RHR System piping.
ANALYSIS OF EVENT
In Mode 5 with Reactor Coolant System loops filled, one RHR loop shall be operable and in operation, and either one additional RHR loop shall be operable or the secondary side water level of at least one steam generator shall be operable pursuant to BVPS Unit 1 Technical Specification 3.4.7. Both trains of RHR were declared inoperable on October 2, 2010 based upon the lack of reasonable assurance that both trains of RHR system remained operable with a cracked weld on 1 RH-200. Condition C of Technical Specification 3.4.7, for no required RHR loops operable, was entered at 1011 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.846855e-4 months <br /> on October 2, 2010 and was not exited until 1156 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.39858e-4 months <br /> on October 6, 2010. With no trains of the RHR System operable in Mode 5, this condition is reportable as an event that could have prevented the fulfillment of a safety function for systems needed to remove decay heat per 10CFR50.73(a)(2)(v)(B).
The Required Action for Condition C of Technical Specification (TS) 3.4.7 requires action immediately to restore one RHR loop to operable status and operation. Since the adverse RHR condition existed when TS 3.4.7 became applicable upon entering Mode 5 at 0453 hours0.00524 days <br />0.126 hours <br />7.490079e-4 weeks <br />1.723665e-4 months <br /> on October 2, 2010 and no action was immediately initiated at 0453 hours0.00524 days <br />0.126 hours <br />7.490079e-4 weeks <br />1.723665e-4 months <br /> (since the significance of the noted leakage was not recognized until 1011 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.846855e-4 months <br />), this was an (inadvertent) operation/condition prohibited by plant's Technical Specifications, and is also reportable per 10 CFR 50.73(a)(2)(i)(B).
The safety significance associated with the cracked weld on the BVPS Unit 1 RHR drain line valve 1 RH-200 is considered to be very low. This is based on the fact that the indicated leakage was insignificant, contingencies were in place to limit any further crack propagation, and in the unlikely event that a pipe rupture occurred, the shutdown safety functions for decay heat removal and RCS inventory control could be maintained.
This event was previously reported as an event that could have prevented the fulfillment of a safety function of systems needed to remove decay heat, pursuant to 10 CFR 50.72(b)(3)(v)(B) at 1538 hours0.0178 days <br />0.427 hours <br />0.00254 weeks <br />5.85209e-4 months <br /> on October 2, 2010 (Event Notification 46304).
CORRECTIVE ACTIONS
- 1. Valve 1 RH-200 was replaced and the pipe length shortened to minimize the susceptibility to natural frequency vibration induced fatigue on October 15, 2010. The new valve was declared operable following required post-maintenance inspection/testing.
- 2. The prior valve 1 RH-200 with a piping segment attached was metallurgically examined to validate the probable cause of the flaw (crack). The results of this examination determined that the mechanism responsible for the cracks in the pipe and socket weld is fatigue.
- 3. The extent of condition at BVPS Unit 1 involved selecting at least two vents/drains for additional examinations during the 1 R20 refueling outage from each of six susceptible systems. PT (penetrant test) and visual examinations were performed on the thirteen selected locations, all with acceptable results.
- 4. The thirteen BVPS Unit 1 potentially susceptible piping locations described in Corrective Action #3 above, along with the 1 RH-200 location, will be evaluated for natural frequency and the potential need for further permanent modifications.
- 5. The engineering Design Interface Review Checklist Form will be revised to ensure that modifications to safety related system vent and drain lines adequately evaluate and minimize susceptibility to vibration induced fatigue weld failures.
- 6. A review will be performed of BVPS Unit No. 2 unsupported RHR System vents/drains along with a sample of other major systems unsupported vents/drains during the upcoming BVPS Unit No. 2 refueling outage scheduled for'the Spring of 2011. Should the sample indicate a design weakness of vibration induced fatigue, an expanded scope may be considered.
- 7. A plant operating experience report was issued on this event on 10/29/2010 (OE 32192).
Completion of the above and other corrective actions are being tracked through the BVPS corrective action program.
PREVIOUS SIMILAR EVENTS
BVPS Unit 1 has not experienced any significant primary coolant leaks from any vent or drain valves since prior to 1994. BVPS Unit 1 Licensee.Event Report 2009-004 described an event involving two notable flaws in a Reactor Coolant System drain line; however these flaws were not through-wall nor did they involve a valve or weld.
BVPS Unit 2 identified a crack in a similar RHR drain valve weld in 2008. A separate condition report has been entered into the Corrective Action Program to evaluate why prior Operating Experience events did not initiate sufficient corrective action to minimize the peobability of the BVPS Unit 1 1RH-200 event occurring in 2010.
CR 10-83533/10-84995