05000259/LER-2018-002, Automatic Reactor Scram Due to an Unanticipated Electro-Hydraulic Control Logic Condition
| ML18137A110 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  (DPR-033) |
| Issue date: | 05/17/2018 |
| From: | Hughes D Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 2018-002-00 | |
| Download: ML18137A110 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 2592018002R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 May 17, 2018 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Subject:
Browns Ferry Nuclear Plant, Unit 1 Renewed Facility Operating License No. DPR-33 NRC Docket No. 50-259 Licensee Event Report 50-259/2018-002-00 10 CFR 50.73 The enclosed Licensee Event Report provides details of an automatic reactor scram due to an unanticipated electro-hydraulic control logic condition. The Tennessee Valley Authority is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(iv)(A), as any event or condition that resulted in a manual or automatic actuation of the RPS and general containment isolation signals affecting containment isolation valves in more than one system or multiple Main Steam Isolation Valves.
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.
D. L. Hughes Site Vice President Enclosure: Licensee Event Report 50-259/2018-002 Automatic Reactor Scram Due to an Unanticipated Electro-Hydraulic Control Logic Condition 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 50-259/2018-002-00 Automatic Reactor Scram Due to an Unanticipated Electro-Hydraulic Control Logic Condition See Enclosed
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
, the NRC may not conduct or sponsor, and a person is not required ID respond ID, the information collection.
3.Page Browns Ferry Nuclear Plant, Unit 1 05000259 1 OF 6
- 4. Title Automatic Reactor Scram Due to an Unanticipated Electro-Hydraulic Control Logic Condition
- 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.
03 18 2018 2018 002 00 05 17 2018 Facility Name Docket Number N/A N/A
- 9. Operating Mode
- 11. This Report is Submitted Pursuant to the Requirements of 10 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) 1 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 5o.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
~ 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 5o.36(c)(2)
D 50.73(a)(2)(v)(B)
D 13.11(a)(5)
D 20.2203(a)(2)(iv)
D 5o.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 13.77(a)(1) 100 D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C)
D OTHER Specify in Abstract below or in
C. Dates and approximate times of occurrences
Date April 8, 2017 January 22, 2018 March 11, 2018 March 11, 2018 March 18, 2018 March 18, 2018 Time (CDT)
Event 0720 1611 1158 1516 EHC megawatt input transducer 2 became non-functional.
EHC megawatt input transducer 1 became non-functional.
Operations personnel received an alarm for Battery Board 2 Breaker Tripout/Fuse Blown, or Ground because Battery Board 2 had a +230 V ground.
A +230 V ground on Battery Board 2 was indicated and identified as a Control Room deficiency.
Unit 1 automatically scrams due to the designed response of the EHC system to a sensed APRM High Flux condition.
Operations personnel reported the scram to the NRC in Event Notification 53269.
D. Manufacturer and model number of each component that failed during the event
No components failed during this event.
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
This reactor scram event was identified through control room indicators and alarms.
G. Failure mode, mechanism, and effect of each failed component No components failed during this event.
H. Operator actions
00 Appropriate operator actions were taken in response to the scram and recovery from the scram.
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-- CT vmtl:.. v. ;jl'1U-V iU4 t:JU"IKt::S : vv* v Estimated burden per r1!sponse to comply with this mandatory col~tion request 80 tours.
Reported lessons learned are incorporated into the licensing prooess and fed bock to induslly. Send commenls regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,0f by e-mail to l nfocol~ls. Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Mairs, NEOB-10202, (3150-0104), Office of Management and Budget Washington, DC 20503. If a means used to if11Xlse an information col~ does oot display a clnen11y valid OMB control number, the NRG may not conduct Of sponSOf, and a person is not required to r1!spond to, the information collection.
YEAR 2018
- 3. LER NUMBER SEQUENTIAL NUMBER 002 REV NO.
00 I.
Automatically and manually initiated safety system responses
The RPS and containment isolation systems automatically responded due to this event, causing Unit 1 to scram and containment isolation.
Ill.
Cause of the Event
A. Cause of each component or system failure or personnel error This event was directly caused by opening and then closing the Generator 1 Relay 250 DC Supply breaker during ground location activities with both Unit 1 megawatt input transducers non-functional.
The root cause investigation determined there was inadequate procedural guidance to assess operational impact(s) for digital logic inputs at the time equipment failures or faults occur (prior to entering the work management process).
No human performance related root cause was identified.
IV.
Analysis of the Event
The Initial Load Pickup Logic uses a combined input from both the Generator load sensors and the Generator Breaker status. With both megawatt input transducers non-functional, a latent condition of "no generator load" input had already existed in the EHC system logic. Closing the Generator 1 Relay 250 V DC Supply breaker in support of identifying the ground location resulted in generating a "Generator Output breaker closed" input. The combination of "no generator load" and "Generator Output breaker closed" signals resulted in the Initial Load Pickup Logic being met. The Initial Load Pickup Logic raises TCV demand in order for the unit to pick-up load when the Generator Breaker is closed. The TCVs open upon Generator Breaker closure to provide more steam to the turbine. The control system imposes a maximum demand for startup conditions, which corresponds to TCVs opened to approximately 10 percent. To satisfy this, the TCVs responded by moving in the Closed direction from their "at-power" positions, causing reactor pressure to rise to the automatic scram set point.
The megawatt input transducers provide defense-in-depth against adverse control system impacts due to Generator Breaker status contact failures. If either of the Generator Power megawatt input transducers into EHC had been functional, then Unit 1 would have remained online.
V.
Assessment of Safety Consequences
This event resulted in the automatic actuation of safety systems. This event did not result in the inoperability or unavailability of any system to provide their required safety functions. All withdrawn
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, the NRC may not conduct or sponsor, and a person is not required to respond to, the information coRection.
YEAR 2018
- 3. LER NUMBER SEQUENTIAL NUMBER 002 REV NO.
00 control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves operating during the initial transient as expected. The Main Turbine Bypass Valves controlled reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals the affected components actuated as required. All safety systems operated as expected. Therefore, this condition had a was of low safety significance and had negligible impact on the health and safety of the public.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event No systems or components failed during this event.
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 This event did not occur when the reactor was shutdown.
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 Safety system availability was not impacted by this event.
VI.
Corrective Actions
This event was entered into the TV A Corrective Action Program and is being tracked under Condition Report 1397341.
A. Immediate Corrective Actions
The immediate corrective action included:
Both megawatt input transducers were replaced prior to Unit 1 restart.
Engineered Solutions, Inc. conducted an independent evaluation for what caused the Unit scram.
A Nuclear Operating Experience Report (NOER) was issued for communication to other TVA sites. This established that Operational Risk Review "Red Sheets" are being utilized to evaluate operational risk prior to work activities. Red Sheets are typically used for emergent work inside of T-3 weeks, but are being used until operational impact reviews are established.
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, the NRC may not conduct ()( spons()(, and a person is not required ID respond ID, the information collection.
YEAR 2018
- 3. LER NUMBER SEQUENTIAL NUMBER 002 REV NO.
00 B. Corrective Actions to Prevent Recurrence or to reduce probability of similar events occurring in the future Clarify operator responsibilities for the performance of operational impact review in procedures, including the development of a tool or checklist for operational impact reviews.
VII.
Previous Similar Events at the Same Site
A review of the BFN CAP and Licensee Event Reports for Units 1, 2, and 3 found no instances of reactor scrams similar to this event within the past five years.
VIII.
Additional Information
None.
IX.
Commitments
None. Page _ 6_ of _ 6_