05000296/LER-2018-003, Misadjusted Switch Results in Condition Prohibited by Technical Specifications
| ML18110A360 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/20/2018 |
| From: | Hughes D Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 2018-003-00 | |
| Download: ML18110A360 (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)(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)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(vii), Common Cause Inoperability |
| 2962018003R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 April 20, 2018 A TIN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Subject:
Browns Ferry Nuclear Plant, Unit 3 Renewed Facility Operating License No. DPR-68 NRC Docket No. 50-296 Licensee Event Report 50-296/2018-003-00 10 CFR 50.73 The enclosed Licensee Event Report provides details of the inoperability of a misadjusted switch which cause systems to be inoperable for longer than allowed by plant Technical Specifications (TS). The Tennessee Valley Authority (TVA) is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(i)(B), as any operation or condition which was prohibited by the plant's TS.
There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact J. L. Paul, Nuclear Site Licensing Manager, at (256) 729-2636.
Respectfully,
~~~~A\\
Site Vice President Enclosure: Licensee Event Report 50-296/2018-003 Misadjusted Switch Results in Condition Prohibited by Technical Specifications cc (w/ Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant
ENCLOSURE Browns Ferry Nuclear Plant Unit 3 Licensee Event Report 50-296/2018-003-00 Misadjusted Switch Results in Condition Prohibited by Technical Specifications See Enclosed
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (02-2018)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the infonnation collection.
3.Page Browns Ferry Nuclear Plant, Unit 3 05000296 1 OF 7
- 4. Title Misadjusted Switch Results in Condition Prohibited by Technical Specifications
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved I
Sequential I Rev Facility Name Docket Number Month Day Year Year Month Day Year N/A N/A Number No.
Facility Name Docket Number 02 20 2018 2018 -
003
- - 00 04 20 2018 N/A N/A
- 9. Operating Mode
- 11. This Report is Submitted Pursuant to the Requirements of 1 O CFR §: (Check all that apply)
D 20.2201 (b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 5 D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D so.13(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iv)(A)
D 50. 73(a)(2)(x)
- 10. Power Level D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 13.11 (a)(4)
D 20.2203(a)(2)(iii)
D so.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73. 71 (a)(S)
D 20.2203(a)(2)(iv)
D 5o.46(a)(3)(ii)
D 50. 73(a)(2)(v)(C)
D 13.77(a)(1) 000 D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50. 73(a)(2)(v)(D)
D 73. n(a)(2)(i)
D 20.2203(a)(2)(vi)
[8J 50. 73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 13.77(a)(2)(ii)
D 5o. 73(a)(2)(i)(C)
D OTHER Specify in Abstract below or in B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event There were no structures, systems, or components (SSCs) whose inoperability contributed to this event.
C. Dates and approximate times of occurrences
Date Time 04/30/2014 0212412016 0530 CST 0212512016 2114 CST 03/03/2016 02/20/2018 1616 CST 02/23/2018 2045 CST Event Procedures for implementing post-maintenance testing (PMT) were revised to remove actions for the testing of all contacts on MJ and MH switches.
DG 3C Emergency LAT was last completed satisfactorily.
DG 3C was removed from service for maintenance.
The MJ(52STA) switch associated with Breaker 1832 was replaced. Adequate post-maintenance testing was not performed following this switch replacement.
During performance of a EOG emergency LAT, the 3EC 4kV SD BD was de-energized and DG 3C started. The expected 4kV load sequencing did not occur.
DG 3C was declared operable following switch replacement and completion of the 3EC Load Acceptance Test.
D. Manufacturer and model number of each component that failed during the event
REV NO.
00 The failed component was a 52STA switch in the MJ position of Siemens Horizontal Vacuum Circuit Breaker 3-BKR-211-03EC/010, Breaker 1832, model number 5-3AF-GEH-250-1200-58.
E. Other systems or secondary functions affected
No other systems or secondary functions were affected by this event.
F. Method of discovery of each component or system failure or procedural error
The switch failure was discovered during performance of a DG 3C Emergency LAT when an open contact prevented the DGVA relays from energizing. These relays were to initiate the 4kV load sequence and 480V load shed during a loss of offsite power with ECCS initiation signal.
G. Failure mode, mechanism, and effect of each failed component Troubleshooting revealed that the condition was caused by a misadjusted breaker STA actuator arm, which did not rotate the STA switch enough to fully close contacts 8-8C. The contact 8-8C failure of the MJ(52STA) switch prevented the initiation of the 4kV load sequence and Unit 3 Division 11 480V load shed during a loss of offsite power with ECCS initiation signal. The 3B and 30 CS pumps, the 3B RHR pump, and the B1 RHRSW Pump would not have automatically started during accident conditions with 4kV SD BO 3EC supplied from DG 3C.
H. Operator actions
There were no operator actions associated with this event.
I.
Automatically and manually initiated safety system responses
There were no automatic or manual safety system responses associated with this event. The 3B RHR pump, the 3B and 30 CS pump, and the B1 RHRSW pump failed to actuate.
Ill.
Cause of the Event
A. Cause of each component or system failure or personnel error The SB-1 and SBM switch assemblies are an inadequate design for use as a breaker operated auxiliary switch. They were originally designed to be manual switches intended for a more moderate operation than the fast, forceful blow of a 4KV breaker stationary switch closure.
The breaker STA switch operator arm engages the breaker cubicle linkage which rotates the STA switch stem to operate the switch. This mechanistic method of STA switch operation can introduce misalignment issues, since small changes in breaker STA switch operator arm height or "slack/slop" in the linkage can cause the STA switch to fail to close the contacts. The STA switch assembly has proven to be non-reliable for safety-related functions. A contributing cause was that the PMT requirements for testing all contacts on MJ and MH switches were removed in a 2014 procedure change.
There were no human performance related root causes associated with this event.
IV.
Analysis of the Event
The pumps impacted by the failure to automatically start affect the Unit 3 CS, RHR, and RHRSW systems. The CS system consists of two independent trains that are each driven by two pumps, which spray water onto the top of the fuel assemblies to cool the core and limit fuel cladding temperature in low-pressure accident conditions. The RHR system consists of four heat exchangers
-- * -~ l:IY UMl:I: NU. 31:>U-U1U4 t:JU"IKt:~: --*-,, ___ _
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
YEAR 2018
- 3. LER NUMBER SEQUENTIAL NUMBER 003 REV NO.
00 and four pumps for each unit. The RHR system serves to restore and maintain the coolant inventory in the reactor vessel so that the core is adequately cooled after a loss of coolant accident (LOCA).
The RHR system also provides cooling for the pressure suppression pool. The RHRSW System is a plant-shared system consisting of twelve pumps that serve to remove heat from the primary water of the RHR systems.
BFN, Unit 3, TS 3.3.5.1 requires the CS B and D pumps and the Low Pressure Coolant Injection (LPCI) RHR pump B are declared inoperable when their associated time delay relays are declared inoperable.
BFN, Unit 3, TS 3.5.1 requires when the auto-start functions for the 38 and 3D CS pumps and the 38 LPCI RHR pump were disabled, two low pressure ECCS injection or spray subsystems were inoperable, which requires immediate TS LCO 3.0.3 entry.
BFN, Unit 3, TS 3.7.1 requires one inoperable RHRSW pump to be restored to Operable status within 30 days to avoid entry into a shutdown LCO.
Additionally, since equipment required for a Mode change was inoperable, BFN, Unit 3, was in violation of TS 3.0.4 on March 26, 2016, during startup from a planned refueling outage, and again on January 11, 2018, during startup from an unrelated scam event.
This event resulted in BFN, Unit 3, auto-start function for the 38 and 3D CS pumps, 38 RHR pump, and the 81 RHRSW pump being inoperable for longer than allowed by plant TS. The auto start function of the pumps was only affected during a loss of offsite power. The automatic start functions of the affected pumps would have operated properly under normal or alternate power. The automatic start capability of the other Unit 3 CS and RHR Pumps, and the remaining 11 common RHRSW Pumps were unaffected by this condition. Additionally, Unit 3 Division I 480V load shed and the remaining 7 DGs were unaffected by this condition. The automatic start function of redundant CS, RHR, and RHRSW trains were unaffected by this condition.
The manual start functions of these systems were not affected by MJ(52STA) switch failure, and Control Room operators could have manually started these pumps when their failure to automatically start was identified.
V.
Assessment of Safety Consequences
A probabilistic risk assessment (PRA) was performed to evaluate the risk significance on internal events. The increase in Core Damage Frequency (CDF) was determined to be less than 1 E-7, and the Large Early Release Frequency (LERF) was determined to be less than 1 E-8. PRA modeling of the impacts on fire risk caused by this event were determined to increase CDF by 2.2E-7 per year, and increase LERF by 6.5E-08 per year. During the time period of inoperability, there was no significant risk to the health and safety of the public or plant personnel resulting from this event.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event The operability of the 3A and 3C CS pumps; the 3A, 3C, and 3D RHR pumps; and the Unit 3A1, C1, and D1 RHRSW pumps was not affected by this event. Each of these pumps were capable of automatically performing their required safety functions.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident During the time the reactor was shutdown, all affected systems remained available to perform their required safety functions under manual actuation.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service lnoperability of the MJ(52STA) switch was discovered on February 20, 2018, at 1616 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.14888e-4 months <br /> CST.
Operability was restored on February 23, 2018, at 1129 CST, but system operability was not required by plant TS during that period since BFN, Unit 3, was in Mode 5. Further evaluation determined that the inoperability of these systems actually began on February 25, 2016, at 2114 CDT, when the breaker was last demonstrated to be operable.
VI.
Corrective Actions
This event was entered into the TV A Corrective Action Program and is being tracked under Condition Reports (CRs) 1389131, 1389133, and 1390278.
A. Immediate Corrective Actions
The MJ(52STA) switch on Breaker 1832 was replaced.
B. Corrective Actions to Prevent Recurrence or to reduce probability of similar events occurring in the future Until long-term corrective actions are implemented, CR actions include developing instructions to verify proper 52STA switch alignment. ECl-O-OOO-SWZ001, Replacement of Type SB switches; PMT-O-OOO-TST001, Post Maintenance Testing Matrices; and other relevant procedures will be revised to add steps to verify the proper alignments of the 52STA switch and the Breaker 52STA Switch Cam. This will address the apparent cause, and prevent recurrence failures until 52STA switch replacement modifications are complete.
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
YEAR 2018
- 3. LER NUMBER SEQUENTIAL NUMBER 003 REV NO.
00 Issues with STA switches were previously recognized in 1991, during preparation for the Unit 2 restart. Modifications installed a parallel connection in Units 1 and 2 EOG breakers to preclude a failure to actuate. This design change was not performed on Unit 3 prior to the Unit 3 restart in 1995.
These design changes were implemented during U3R 18, and the 52a switch is now utilized to execute safety related functions for the EOG breakers and 4 kV shutdown board normal supply breakers. Similar design changes are planned for future implementation on all BFN, Units 1, 2, and 3, and common breakers that utilize STA switches to execute safety related functions.
VII.
Previous Similar Events at the Same Site
A search of the Corrective Action Program for BFN, Units 1, 2, and 3, identified eight MJ(52STA) switch failure events since 2010. These failures were captured by CRs 230836, 328038, 672598, 752488, 792179, 801449, 980277, and 1140776 (associated with LER 296/2016-002-01).
These individual failures were collectively evaluated by CR 803629 described below. CR 803629 was written in June 2014 to document the trend of 4 kV breaker's MJ(52STA) stationary contact failures, the same failure that resulted in this event. The cause evaluation for CR 803629 identified two apparent causes:
- 1. An engineering evaluation concluded that switches should only be inspected for failure, since this was determined to be more conservative than the vendor-recommended periodic cleaning and inspection strategy.
- 2. The breaker support components were overlooked with respect to reliability despite being a vital component to the reliability of the breaker.
Previous corrective actions have proven ineffective at permanently resolving the repetitive failure of the MJ(52STA) switches.
VIII.
Additional Information
None.
IX.
Commitments
None. Page _7_ of _7_