05000259/LER-2010-003, Regarding Failure of a Low Pressure Coolant Injection Flow Control Valve
| ML103620327 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/22/2010 |
| From: | Polson K Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 10-003-00 | |
| Download: ML103620327 (9) | |
| 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 |
| 2592010003R00 - NRC Website | |
text
Tennessee Valley Authority; Post Office Box 2000, Decatur, Alabama 35609-2000 December 22, 2010 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D. C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Facility Operating License No. DPR-33 NRC Docket No. 50-259
Subject:
Licensee Event Report 50-25912010-003, Revision 0 The enclosed Licensee Event Report (LER) provides details of a failure of a low pressure coolant injection flow control valve and the resulting failure to meet the requirements of Browns Ferry Nuclear Plant, Unit 1 Technical Specification 3.5.1 concerning low pressure coolant injection operability.
The Tennessee Valley Authority (TVA) 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.
TVA is' currently completing the investigation and evaluation for this event. Upon completion of these actions, TVA will submit a revised LER.
There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact J. E. Emens, Jr., Site Nuclear Licensing Manager, at (256) 729-2636.
Respectfully, K.
olson Vice President Enclosure: Licensee Event Report - Failure of a Low Pressure Coolant Injection Flow Control Valve
U.S. Nuclear Regulatory Commission Page 2 December 22, 2011 cc (w/ Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant
ENCLOSURE Browns Ferry Nuclear Plant Unit 1 Licensee Event Report - Failure of a Low Pressure Coolant Injection Flow Control Valve SEE ATTACHED
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/13/2013 (10-2010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Browns Ferry Nuclear Plant Unit 1 05000259 1 OF6
- 4. TITLE Failure of a Low Pressure Coolant Injection Flow Control Valve
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER MO YEAR YEAR NUMBER NO.
N/A 05000 FACILITY NAME DOCKET NUMBER 10 23 2010 2010-003 -00 12 22 2010 N/A 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
E] 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[1 50.73(a)(2)(vii) 3 [1 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
E] 20.2203(a)(4)
[] 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 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 El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
[] 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4)
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
[E 73.71(a)(5) 0 El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)
[E OTHER S20.2203(a)(2)(vi)
ED 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in C. Dates and ADDroximate Times of Maior Occurrences:
w.
June 2006 March 13, 2009, at 0553 hours0.0064 days <br />0.154 hours <br />9.143518e-4 weeks <br />2.104165e-4 months <br /> October 23, 2010, at 1417 hours0.0164 days <br />0.394 hours <br />0.00234 weeks <br />5.391685e-4 months <br /> October 23, 2010, at 1433 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.452565e-4 months <br /> October 23, 2010, at 1505 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.726525e-4 months <br /> LPCI flow control valve 1 -FCV-074-066 was refurbished prior to the Unit 1 restart. A previous design change to reinforce the integrity of the valve internals was not incorporated.
During the Unit 1 Cycle 7 forced outage, RHR Loop II was in service for Shutdown Cooling.
When Shutdown Cooling was secured per RHR System Operating Instruction (01) 1-01-74, 1-FCV-074-066 was closed (this was the last confirmed successful operation of the valve).
During the Unit 1 Cycle 8 refueling outage, Operations personnel attempted to place RHR Loop II in service for Shutdown Cooling in accordance with 1-01-74. Flow could not be confirmed. Problem Evaluation Report (PER) 271338 was subsequently initiated.
Operations personnel placed RHR Loop I in shutdown cooling in accordance with 1-01-74.
Unit 1 entered Mode 4.
D. Other Systems or Secondary Functions Affected
Analysis of other systems or secondary functions affected is still in progress, and will be provided in a supplement to this LER.
E. Method of Discovery
The valve failure was discovered during the performance of 1-01-74, "Residual Heat Removal System," Section 8.12.2, "Initiation/Operation of RHR Loop II in Shutdown Cooling."
F. Operator Actions
None G. Safety System Responses:
None Ill.
CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause for this condition was a LPCI flow control valve disc-stem separation, with the valve disc stuck in the closed position.
NRC FORM 366 (10-2010)
I
B. Root Cause:
TVA is currently completing the root cause analysis for this event; therefore, the following information is considered preliminary. Valve 1-FCV-074-066 is physically located in the RHR LPCI Loop II flow path and is normally in the open position for the passive safety-related function of the valve. The valve is used to throttle shutdown cooling flow and is closed to divert flow from the LPCI flow path when containment cooling is desired. There are six of these type valves, two per unit with one in each loop of LPCI.
In 1975, Engineering Change Notices (ECNs) L1473 (BFN 1 and 2) and L1217 (BFN 3) were developed to implement a disc modification to eliminate excessive vibration experienced at low flow and high pressure drop conditions. This modification affected one valve per unit and removed the existing discs from the 24-inch pressure seal angle valves and replaced them with V-notch discs supplied by the original equipment manufacturer, the Walworth Company. In addition, a disc locking key was placed through the skirt into the keyway machined into the stem to prevent independent rotation between skirt, disc, and stem. This change was performed to support a test to ascertain if this modification would be successful in reducing vibration while maintaining the system performance characteristics and to assist in controlling the reactor cool-down rate. In 1978 ECN L2107 was issued to implement the change on the remaining three valves on all three units.
Valve 1 -FCV-074-066 was refurbished in 2006 prior to the restart of Unit 1 after an extended outage. The design change noted in the ECNs above was not incorporated into this valve, apparently because the ECNs failed to update vendor drawings and provide detail for the field modifications. Therefore, the preliminary root cause of this event was inadequate configuration control.
IV.
ANALYSIS OF THE EVENT
TVA is currently completing the analysis for this event; therefore, the following information is considered preliminary. The condition being reported is the operation of Unit 1 in a manner prohibited by TS.
1 -FCV-074-066 is normally in the open position for the passive safety-related function of the valve. The valve is used to throttle shutdown cooling flow and is closed to divert flow from the LPCI flow path when containment cooling is desired. The valve was found to be failed closed while attempting to establish shutdown cooling during the Unit 1 Cycle 8 refueling outage. Operations personnel secured RHR Loop II shutdown cooling and established shutdown cooling using RHR Loop I in accordance with 1-01-74.
Following preliminary investigations, the failed valve was reworked and tested, and RHR Loop II was returned to service.
Based on preliminary causal information, the possible extent of condition was determined to be limited to the corresponding valve in RHR Loop I of Unit 1, and to the corresponding valves in RHR Loops I and II of Units 2 and 3. Boroscope inspections were performed on the Unit 1 Loop I LPCI flow control valve 1-FCV-074-052 during the refueling outage. The results revealed that the valve internals were properly modified. No signs of fatigue or damage were found. These corresponding valves and associated LPCI RHR Loops on Units 2 and 3 are considered to be operable based upon successful shutdown cooling 1
NRC FORM 366 (10-2010)
operation during recent outages and/or monthly venting, which confirm that the valves are in the open, safety-related position. In addition, the valves on Units 2 and 3 have been in service longer than the failed Unit 1 valve.
V.
ASSESSMENT OF SAFETY CONSEQUENCES
TVA is currently completing the analysis for this event; therefore, the following information is considered preliminary. The applicable safety-related mode for the RHR system is to provide a flow path for transmission of water supply to the reactor for a mission time up to 30 days for core cooling following initiation. Unit 1 TS LCO 3.5.1 requires both RHR loops of LPCI to be operable for ECCS in reactor Modes 1, 2, and 3. Unit 1 appears to have operated with RHR Loop II inoperable longer than allowed by TS LCO 3.5.1.
Past operability risk analyses associated with this event are ongoing. A specific time of valve failure (i.e., RHR Loop II of LPCI inoperable for ECCS) has not yet been estimated by the root cause analysis investigations. Preliminary indications are that some time between March 13, 2009 (last confirmed successful operation of the valve), and October 23, 2010, could be the most probable time of failure. Actions have been taken to obtain forensic metallurgic testing to assist in this determination; however, this information is not yet available.
Therefore, the assessment of the safety consequences of this event is ongoing and pending the completion of the causal analysis and estimation of the valve failure time.
VI.
CORRECTIVE ACTIONS
TVA is currently completing the root cause analysis and developing associated corrective actions for this event; therefore, the following corrective actions may be revised or supplemented in a later revision to this LER.
A.
Immediate Corrective Actions
- 1. Valve 1 -FCV-074-066 was repaired and tested as operable.
- 2. To address extent of condition of the similar valve on Unit 1 RHR Loop 1, boroscope inspections were performed on Loop I LPCI flow control valve, 1 -FCV-074-052, during the refueling outage. The results revealed that the valve internals were properly modified. No signs of fatigue or damage were found.
- 3. The removed skirt, yoke nut, and stem for valve 1 -FCV-074-066 were sent to Westinghouse for metallurgical analysis to validate suspected failure modes and estimate the time of failure. Southwest Research Institute has taken replication samples of the welds for additional verification for suspected failure modes.
B.
Corrective Actions to Prevent Recurrence:
An administrative change was issued to update appropriate design documents associated with ECNs L1473, L1217, and L2107 (Problem Evaluation Report (PER) 279911).
C.
Additional Corrective Actions
Ultrasonic examination of the corresponding valves in Units 2 and 3 has been scheduled for the period December 29, 2010, through January 14, 2011. This will provide additional confirmation that these valves have not experienced the same failure as valve 1-FCV-074-066.
NRCJ FORIM 366t (10-2010)
VII.
ADDITIONAL INFORMATION
A.
Failed Components:
The RHR Loop II LPCI flow control valve, 1-FCV-074-066, was manufactured by the Walworth Company as Part No. 531,543, Drawing No. A-12337-M-1. The valve is a 24-inch cast carbon steel, butt welded, pressure-seal angle globe valve operated by a Limitorque SMB-5T-350 motor operator.
B.
Previous LERS or Similar Events:
This information is still under investigation.
C.
Additional Information
The corrective action document for this report is PER 271338. PER 279911 is an associated corrective action document used to document and track the update of the design output of the approved design changes.
D.
Safety System Functional Failure Consideration:
Pending completion of the assessment of the safety consequences, the preliminary conclusion is that this event is not a safety system functional failure according to NEI 99-02.
E.
Scram With Complications Consideration:
This event did not include a reactor scram.
VIII.
COMMITMENTS
None NRC FORM 366 (10-2010)