05000410/LER-1917-001, Regarding Automatic Reactor Scram Due to High Reactor Pressure
| ML17296A158 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/04/2017 |
| From: | Kreider R Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP2L 2655 LER 17-001-00 | |
| Download: ML17296A158 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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) |
| 4101917001R00 - NRC Website | |
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- Exelon Generation NMP2L 2655 October 4, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 Docket No. 50-410 10 CFR 50.73
Subject:
NMP2 Licensee Event Report 2017-001, Automatic Reactor Scram due to High Reactor Pressure In accordance with the reporting requirements contained in 10 CFR 50.73(a)(2)(iv)(A), please find enclosed NMP2 Licensee Event Report (LER) 2017-001, Automatic Reactor Scram due to High Reactor Pressure.
There are no regulatory commitments contained in this letter.
Should you have any questions regarding the information in this submittal, please contact Dennis M. Moore, Site Regulatory Assurance Manager, at (315) 349-5219.
r;Jly, *<~-
lbert E. Kreider Jr.
Plant Manager, Nine Mile Point Nuclear Station Exelon Generation Company, LLC REK/KJK
Enclosure:
NMP2 Licensee Event Report 2017-001, Automatic Reactor Scram due to High Reactor Pressure cc:
NRC Regional Administrator, Region I NRC Resident Inspector NRC Project Manager
f' Enclosure NM P2 Licensee Event Report 2017-001, Automatic Reactor Scram due to High Reactor Pressure Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69
NRC FORM 366 U.S.. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017)
, the h!tQ:l/www.nrc.govlreading-rmldoc-collectionslnuregslstafflsr10221r3l)
NRG may not conduct or spo.nsor, and a person is not required to respond to, the information collection.
- 3. PAGE Nine Mile Point Unit 2 05000410 1 OF 5
- 4. TITLE Automatic Reactor Scram due to High Reactor Pressure
- 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 NUMBER NO.
MONTH DAY YEAR NIA NIA FACILITY NAME DOCKET NUMBER 08 05 2017 2017 - 001
- - 00 10 04 2017 NIA NIA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.2201 (bl D 20.2203(a)(3)(il 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)(1l D 20.2203(a)(4)
D so.13(a)(2l(iiil D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(il D
50.36(c)(1)(i)(A) r8] 50.73(a}(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(iil D
50.36(c)(1)(ii)(A)
D 50.73(a)(2)(v)(A)"
D 13.11 (a)(4l D 20.2203(a)(2)(iiil D
so.36(c)(2l D
50.73(a)(2)(v)(B)
D 13.11(a)(s)
D 20.2203(a)(2)(iv)
D so.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 13.77(a)(1) 084 D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 13.11(a)(2)(il D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 13.11(a)(2)(iil D
50.73(a)(2)(i)(C)
D OTHER Specify in Abstract below or in D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES AND OPERA TOR ACTIONS:
The dates, times, and major occurrences and operator actions for this event are as follows. All dates are 2017.
August 5 1500 Began power ascension from 20% after completion of turbine controls troubleshooting and overspeed trip repairs.
2040 Power ascension placed on hold at 84% power and started TSV and TCV testing per N2-0SP-RPS-Q001.
2235 During step 8.8.3, when testing the last set of TSVs, the TCVs started closing, the TBVs opened and the reactor scrammed on high reactor pressure. EOP-RPV was entered due to low reactor water level, as well as high reactor pressure. N2-SOP-101 C was entered due to the reactor scram.
2323 Exited EOP-RPV and N2-SOP-101C.
E. METHOD OF DISCOVERY
This event was self-revealing during the performance of the surveillance test when relays were heard actuating unexpectedly and a reactor scram occurred.
F. SAFETY SYSTEM RESPONSES:
REV NO.
00 All systems operated as expected. After the scram, there were no safety system actuations.
II. CAUSE OF EVENT
The definitive root cause of the equipment failure was not located but was bound to spurious actuation of load limit relays KL 186 and KL 187. The spurious action was caused by an intermittent ground and/or an induced voltage within the load limit circuit. This is a result of the non-fault tolerant original design of the Electro-hydraulic Control (EHC) system.
III. ANALYSIS OF THE EVENT
The automatic reactor scram is reportable under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR
- 50. 73(a)(2)(iv)(A). It is defined under paragraph 10 CFR 50. 73(a)(2)(iv)(A) as any event or condition that resulted in manual or automatic action of any of the systems listed in 10 CFR
- 50. 73(a)(2)(iv)(B).
The definitive root cause of the equipment failure was not located but was bound to spurious actuation of load limit relays KL 186 and KL 187. The spurious action was caused by an intermittent ground and/or an induced voltage within the load limit circuit. The relay actuation resulted in the TCVs closing, the TBVs opening and a reactor scram on high reactor pressure.
All control rods fully inserted and there were no safety system actuations.
Based on the above discussion, it is concluded that the safety significance of this event is low and the event did not pose a threat to the health and safety of the public or plant personnel.
Troubleshooting identified an erroneous control valve demand signal resulting in fast closure of the TCVs and subsequent increase in reactor pressure. After instrumenting the load control circuit, an anomaly on the load limit amplifier card was identified. The load limit amplifier card was replaced. Failure analysis did not identify any anomalies with the card.
To further protect against latent failures, leads have been lifted on the stop valve load limit switches to prevent fast closure of the TCVs in the event of a stuck load limit switch during future valve testing.
This event does affect the NRC Regulatory Oversight Process Indicator for unplanned scrams per 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> of critical operation.
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
The load limit amplifier card was replaced.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
To prevent recurrence, replacement of the current obsolete and single point vulnerable NMP2 Turbine EHC system with a fault tolerant Digital EHC system is scheduled for the next refueling outage.
To further protect against latent failures, a temporary modification to lift the leads on the stop valve load limit switches was implemented. Lifting the leads minimizes the risk of fast closure of the control valves in the event of a stuck load limit switch during future turbine valve testing. Additionally, to minimize risk prior to replacement of the system, the testing method for TSVs is being revised prior to the next performance.
REV NO.
00
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
The definitive root cause of the equipment failure was not located but was bound to spurious actuation of load limit relays KL 186 and KL 187. The spurious action was caused by an intermittent ground and/or an induced voltage within the load limit circuit.
The load limit amplifier card was replaced, but troubleshooting determined it was not the single cause of the event. Potentially stuck limit switches have been disabled.
B. PREVIOUS LERs ON SIMILAR EVENTS:
None C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
REV NO.
00 COMPONENT IEEE 803 'FUNCTION IDENTIFIER IEEE 805 SYSTEM IDENTIFICATION Electro Hydraulic Control System Load Limit Amplifier Card Load Limit Relay Limit Switch Turbine Stop Valve Turbine Control Valve Turbine Bypass Valve Main Turbine Reactor Protection System D. SPECIAL COMMENTS:
None N/A CBD RLY 33 ISV ISV PCV TRB N/A TG TG TG SB SB SB SB TA SC Page _s_ of _s_