05000259/LER-2012-006, For Browns Ferry Nuclear Plan, Unit 1, Regarding High Pressure Coolant Injection System Turbine Failed to Trip Using the Manual Trip Pushbutton
| ML12172A395 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/18/2012 |
| From: | Polson K Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 12-006-00 | |
| Download: ML12172A395 (10) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(vii)(A), Common Cause Inoperability 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), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2592012006R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 June 18, 2012 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 50-25912012-006-00 The enclosed Licensee Event Report provides details of the High Pressure Coolant Injection System turbine failure to trip using the manual trip pushbutton. The Tennessee Valley Authority is submitting this report in accordance with 10 CFR 50.73(a)(2)(v)(B) and (D), as any event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to remove residual heat and mitigate the consequences of an accident.
The causal analysis for this event is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this 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., Nuclear Site Licensing Manager, at (256) 729-2636.
Re ec ully, K J. Polso V i ePresid
,dK
U. S. Nuclear Regulatory Commission Page 2 June 18, 2012
Enclosure:
Licensee Event Report 50-259/2012-006 High Pressure Coolant Injection System Turbine Failed to Trip Using the Manual Trip Pushbutton cc (w/ Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant
ENCLOSURE Browns Ferry Nuclear Plant Unit I Licensee Event Report 50-25912012-006-00 High Pressure Coolant Injection System Turbine Failed to Trip Using the Manual Trip Pushbutton See Attached
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/31/2013 (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 of 7
- 4. TITLE: High Pressure Coolant Injection System Turbine Failed to Trip Using the Manual Trip Pushbutton
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO.
N/A 05000 FACILITY NAME DOCKET NUMBER 04 19 2012 2012 -
006 00 06 18 2012 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)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[:1 50.73(a)(2)(vii) 0l 20.2201(d)
E0 20.2203(a)(3)(i)
[l 50.73(a)(2)(i)(A)
El 50.73(a)(2)(vii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B) 1E 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
[0 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
[P L 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) 0l 50.46(a)(3)(ii) 50.73(a)(2)(v)(B)
[1 73.71(a)(5) 100 [E
20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C) 0l OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D) speForn Abrtmct rinNRC
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Eric Bates, Licensing Engineer 256-614-7180CAUSE SYSTEM COMPONENT MANU-REPORTABLE MANU-REPORTABLE FACTURER TO EPIX
CAUSE
FACTURER TO EPIX X
BJ FCV S075 Y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED SUBMISSION MONTH DAY YEAR 0
YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
El NO DATE 07 18 2012 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On April 19, 2012, at approximately 1117 hours0.0129 days <br />0.31 hours <br />0.00185 weeks <br />4.250185e-4 months <br /> Central Daylight Time (CDT), the Browns Ferry Nuclear Plant, Unit 1, High Pressure Coolant Injection (HPCI) System was declared inoperable for performance of surveillance procedure 1-SR-3.5.1.7, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor Pressure. On April 19, 2012, at approximately 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br /> CDT, at the end of the HPCI pump run for surveillance procedure 1-SR-3.5.1.7, while attempting to trip the HPCI turbine using manual trip pushbutton 1-HS-073-0018A, the HPCI turbine failed to trip. The manual trip pushbutton was pressed four times with no results. Pressing the manual trip pushbutton should have resulted in the closure of the HPCI turbine stop valve 1 -FCV-073-0018. As a result of the failure, the capability of electrical and mechanical automatic and manual trip functions of the HPCI System was lost.
The HPCI turbine was stopped by securing the HPCI steam supply with closure of the HPCI turbine steam admission valve 1-FCV-073-0016.
The causal analysis for this event is ongoing. Upon completion of the causal analysis, the Tennessee Valley Authority will submit a supplement to this Licensee Event Report.
NRC FORM 366 (10-2010)
PLANT CONDITION(S)
At the time of the event, Browns Ferry Nuclear Plant (BFN), Unit 1, was in Mode 1 at approximately 100 percent rated thermal power.
I1.
DESCRIPTION OF EVENT
A. Event On April 19, 2012, at approximately 1117 hours0.0129 days <br />0.31 hours <br />0.00185 weeks <br />4.250185e-4 months <br /> Central Daylight Time (CDT), the BFN, Unit 1, High Pressure Coolant Injection (HPCI) [BJ] System was declared inoperable for the performance of surveillance procedure 1-SR-3.5.1.7, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor Pressure. During the prerequisite steps of surveillance procedure 1-SR-3.5.1.7, the HPCI turbine stop valve [FCV] 1-FCV-073-0018 was tested and performed as expected. On April 19, 2012, at approximately 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br /> CDT, at the end of the HPCI pump run for surveillance procedure 1-SR-3.5.1.7, while attempting to trip the HPCI turbine using the manual trip pushbutton 1-HS-073-001 8A, the HPCI turbine failed to trip. The manual trip pushbutton 1-HS-073-0018A was pressed four times with no results. Pressing the manual trip pushbutton should have resulted in the closure of the HPCI turbine stop valve 1-FCV-073-0018. As a result of the failure, the capability of electrical and mechanical automatic and manual trip functions of the HPCI System was lost. The HPCI turbine was stopped by securing the HPCI steam supply with the closure of the HPCI turbine steam admission valve 1-FCV-073-0016.
The HPCI turbine stop valve was inoperable until April 24, 2012, when WO 113426235 was performed to rebuild the valve using a different valve disc and cover and post maintenance testing was completed.
B. Inoperable Structures, Components, or Systems that Contributed to the Event The inoperable component that contributed to this event was the HPCI turbine stop valve 1-FCV-073-0018.
C. Dates and Approximate Times of Major Occurrences
April 19, 2012, at 1117 hours0.0129 days <br />0.31 hours <br />0.00185 weeks <br />4.250185e-4 months <br /> CDT HPCI System declared inoperable for performance of surveillance procedure 1-SR-3.5.1.7.
April 19, 2012, at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br /> CDT During the performance of surveillance 1-SR-3.5.1.7, the HPCI turbine stop valve did not close when the manual trip pushbutton was pressed four times.
April 24, 2012, at 1940 hours0.0225 days <br />0.539 hours <br />0.00321 weeks <br />7.3817e-4 months <br /> CDT HPCI System declared Operable after the HPCI turbine stop valve was rebuilt and post maintenance testing was completed.
D. Other Systems or Secondary Functions Affected
There were no other systems or secondary functions affected for this event.
E. Method of Discovery
This event was discovered during the performance of 1-SR-3.5.1.7 by Operations personnel when attempting to manually trip the HPCI turbine.
F. Operator Actions
Operations personnel stopped the HPCI turbine by securing the HPCI steam supply with the closure of the HPCI turbine steam admission valve 1-FCV-073-0016.
G. Safety System Responses There were no safety system responses for this event.
Ill.
CAUSE OF THE EVENT
A. Immediate Cause The immediate cause was the gap between the piston rod of the actuator stem and pilot valve with stem (valve stem) on the HPCI turbine stop valve 1-FCV-073-0018 being set inappropriately during the performance of mechanical corrective instruction MCI-0-073-VLVO01, High Pressure Coolant Injection Turbine Stop Valve - FCV 73-18 Disassembly, Inspection, Rework and Reassembly, in 2008.
B. Root Cause The causal analysis for this event is ongoing. Upon completion of the causal analysis, the Tennessee Valley Authority (IVA) will submit a supplement to this Licensee Event Report (LER) with the root cause.
C. Contributing Factors The causal analysis for this event is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this LER with the contributing factors.
IV. ANALYSIS OF THE EVENT
The causal analysis for this event is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this LER.
TVA is submitting this report in accordance with 10 CFR 50.73(a)(2)(v)(B) and (D), as any event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to remove residual heat and mitigate the consequences of an accident.
On March 23, 2004, MCI-0-073-VLVO01, High Pressure Coolant Injection Turbine Stop Valve - FCV 73-18 Disassembly, Inspection, Rework and Reassembly, Revision 8, was issued with a change made to the amount of clearance for BFN, Unit 3, that is to be set between the valve stem and the piston rod of the actuator. This change occurred to compensate for a valve stem that was acquired from another utility which was believed to be approximately 3/16" shorter than the valve stem that was removed. Prior to this
revision, all three BFN units had a common gap setting guidance of 1/32" to 1/16" per the Electrical Power Research Institute (EPRI) guidelines and General Electric (GE)
Service Information Letter (SIL) 306 Revision 1. Therefore, this revision to MCI-0-073-VLVO01 changed a unit common procedure to specify a BFN, Unit 2, gap and a BFN, Unit 3, gap between the valve stem and piston rod of the actuator. Since BFN, Unit 1, was shut down at the time, no guidance was added to MCI-0-073-VLVO01 to clarify how much of a gap between the valve stem and piston rod of the actuator should be included during reassembly of the BFN, Unit 1, HPCI turbine stop valve. BFN, Unit 1, was restarted in May 2007, three years after the difference in stem gap settings was inserted into MCI-0-073-VLVO01. MCI-0-073-VLVO01 was listed for review during BFN, Unit 1, restart activities, but no revision was made to incorporate BFN, Unit 1, guidance.
Despite the lack of BFN, Unit 1, guidance in MCI-0-073-VLVO01 with regard to the size of the gap between the valve stem and piston rod of the actuator, personnel that performed the work during the BFN, Unit 1, restart set the gap at 0.040". The proper gap is between 1/32" and 1/16" (0.03125"- 0.0625"). Thus, from the time that BFN, Unit 1, was restarted, until the first BFN, Unit 1, refueling outage after restart in 2008, the HPCI turbine stop valve stem assembly was installed with the appropriate clearances.
On November 19, 2008, during the BFN, Unit 1, refueling outage, the piston rings on the HPCI turbine stop valve actuator were replaced due to its failure of the seal leak test performed during the refueling outage. The stems were separated to facilitate piston ring replacement. During reassembly, when the piston was relaxed in its lowest position, the gap between the valve stem and the piston rod was 5/8". The procedure being used did not specify the gap that should be used for BFN, Unit 1, since it was written as a unit common procedure, but instead only listed BFN, Units 2 and 3, gap information. As a result, Maintenance personnel signed the procedure step that required the use of a jack to properly position the piston, yet installed the split coupling with the gap at 5/8". Since the recommended setting is 1/32" to 1/16", the as-left measurement during this evolution was approximately 9/16" greater than it should have been. When the HPCI turbine stop valve strokes open, the disc travels upward into the guide of the top cover. However, increasing the length of the valve stroke by this amount pushes the HPCI turbine stop valve disc into a section of the top cover guide that narrows as it slopes inward. During the time period between November 19, 2008, and April 19, 2012 (when the valve failed to close), more than 350 valve strokes occurred with this extended stem length. The majority of these valve strokes were in support of system and maintenance testing, including oil sampling. With each stroke, the disc and cover guide wore against each other increasing the wear on each component until the friction between the two became greater than the force supplied by the closure spring inside the actuator and the valve failed to close.
Extent of Condition The causal analysis for this event is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this LER.
Extent of Cause The causal analysis for this event is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this LER.
V.
ASSESSMENT OF SAFETY CONSEQUENCES
The HPCI System permits the nuclear plant to be shut down while maintaining sufficient reactor vessel water inventory until the reactor vessel is depressurized. The HPCI System continues to operate until the reactor vessel pressure is below the pressure at which Low Pressure Coolant Injection (LPCI) [BO] operation or Core Spray System operation maintains core cooling. If a Loss of Coolant Accident (LOCA) occurs, the reactor scrams upon receipt of a low-water-level signal or a high-drywell-pressure signal.
HPCI starts when the water level reaches a preselected height above the core, or if high pressure exists in the primary containment (drywell). HPCI should automatically stop when a high water level in the reactor vessel is signaled. With the HPCI turbine stop valve bound open, this could not have occurred. After a trip, the HPCI System is supposed to automatically reset and restart if vessel water level again reaches the pre-selected height above the core. This restart capability would have been impacted by the inability to automatically triplreset and the difficulty in restarting the system through other means.
With the BFN, Unit 1, HPCI turbine stop valve bound in the open position, electrical and mechanical HPCI trip functions were effectively disabled. In this condition, when the HPCI turbine reaches an overspeed condition and does not trip, a failure of the HPCI turbine could result. In addition, due to the failure of the HPCI turbine stop valve to close, HPCI would not trip on high water level. If manual operator action were not taken in this event, HPCI would fill the steam lines with water. This would subsequently send water to the Main Turbine [TRB], the Reactor Core Isolation Cooling (RCIC) System [BN]
steam supply, the Main Feedwater System steam supply, and to the HPCI System steam supply lines, thereby compromising the ability of these systems to mitigate postulated accidents and transients.
Despite this reduction in defense-in-depth, redundant systems such as the Automatic Depressurization System (ADS) [SB], the Core Spray (CS) [BM] System, and the Residual Heat Removal System (RHR) [BO] remained Operable, as allowed by Technical Specifications (TS), to respond to postulated accidents and maintain safe shutdown capability. The Emergency Core Cooling System (ECCS) subsystems (including ADS) are designed to ensure, in the event of a design basis accident and a worst case single failure, that adequate core cooling is maintained. The redundant capacity of the ECCS is consistent with the assumptions used in the safety analyses.
With one low pressure ECCS subsystem inoperable and assuming the HPCI System is inoperable, adequate core cooling is ensured by the operability of the ADS and the remaining low pressure ECCS subsystems. BFN, Unit 1, TS provide an action completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the HPCI System inoperable and one low pressure ECCS injection/spray (either LPCI or CS) subsystem inoperable. During the time period that the incorrect gap setting existed in the HPCI turbine stop valve, a low pressure ECCS subsystem was not concurrently inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In addition,
the remaining ECCS subsystems (including ADS) remained Operable during the time period that the incorrect gap setting existed in the HPCI turbine stop valve, except for approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on March 30, 2011, when an inadvertent actuation of the Common Accident Signal (CAS) Logic of BFN, Unit 2, occurred. This actuation of the CAS Logic caused the inoperability of the BFN, Unit 1, Loop II LPCI and CS subsystems. As a result of the inoperability of the Loop II LPCI and CS subsystems BFN, Unit 1, immediately entered Limiting Condition of Operation (LCO) 3.0.3 as required by TS. LCO 3.0.3 requires action to be initiated within one hour to place the unit in Mode 2 within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, Mode 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, and Mode 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
In addition, during the time period that the incorrect gap setting existed with the HPCI turbine stop valve, the RCIC System was not concurrently inoperable for longer than the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown action completion time provided by the BFN Unit 1 TS, except for the following periods when the RCIC system was inoperable for maintenance.
" Approximately 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> from February 5, 2010, to February 7, 2010 Approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> from August 30, 2010, to August 31, 2010 Approximately 8 days from March 14, 2011, to March 22, 2011 The causal analysis for this event, including additional analysis of the safety consequences of this event, is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this LER with the results of the additional analysis of safety consequences.
VI.
CORRECTIVE ACTIONS - The corrective actions are being managed by TVAs corrective action program.
A. Immediate Corrective Actions
- 2. Verified that the correct gap is set on BFN Unit 2 HPCI turbine stop valve and BFN Unit 3 HPCI turbine stop valve.
B. Corrective Actions to Prevent Recurrence The causal analysis for this event is ongoing. Upon completion of the causal analysis, TVA will submit a supplement to this LER with the corrective actions to prevent recurrence.
VII.
ADDITIONAL INFORMATION
A. Failed Components The failed component was 1-FCV-073-0018. This component was manufactured by Schutte and Koerting Co. with a manufacturer model number of J53MTR.
B. Previous Similar Events
A search of LERs for BFN Units 1, 2, and 3, for approximately the past five years, did not identify any similar events.
Similar Problem Evaluation Reports (PERs) associated with this condition are documented in PERs 377771 and 377833.
C. Additional Information
The corrective action document for this report is PER 539040.
D. Safety System Functional Failure Consideration In accordance with NEI 99-02, this event is considered a safety system functional failure. This event is considered a safety system functional failure because it could have prevented fulfillment of the HPCI System safety functions to remove residual heat and to mitigate the consequences of an accident.
E. Scram With Complications Consideration This event did not include a reactor scram.
VIII. COMMITMENTS
There are no commitments.