05000220/LER-2019-001, Automatic Reactor Scram Due to High Reactor Pressure
| ML19169A058 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/13/2019 |
| From: | Kreider R Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP1L3290 LER 2019-001-00 | |
| Download: ML19169A058 (6) | |
| 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)(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) 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 2202019001R00 - NRC Website | |
text
1 Exelon Generation NMP1L3290 June 13, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 Docket No. 50-220 10 CFR 50.73
- Subject:
NMP1 Licensee Event Report 2019-001, Automatic Reactor Scram Due to High Reactor Pressure In accordance with the reporting requirements contained in 1 O CFR 50.73(a)(2)(iv)(A), please find enclosed NMP1 Licensee Event Report (LER) 2019-001, 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 Brandon Shultz, Site Regulatory Assurance Manager, at (315) 349-7012.
Respectfully, Robert E. Kreider Jr.
Plant Manager, Nine Mile Point Nuclear Station Exelon Generation Company, LLC REK/KJK
Enclosure:
NMP1 Licensee Event Report 2019-001, Automatic Reactor Scram Due to High Reactor Pressure.
cc:
NRG Regional Administrator, Region I NRG Resident Inspector NRG Project Manager
Enclosure NMP1 Licensee Event 2019-001, Automatic Reactor Scram Due to High Reactor Pressure Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63
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, the httQ ://www. n re. gov/reading-rm/doc-col I ections/n u regs/staff/sr1 022/r3/)
NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Nine Mile Point Unit 1 05000220 1 OF 4
- 4. TITLE Automatic Reactor Scram due to High Reactor Pressure
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 04 14 2019 2019 - 001
- - 00 06 13 2019 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 (bl D 20.2203(a)(3)(il D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
N D 20.2201 (dl D 20.2203(a)(3l(iil D 50.73(a)(2)(ii)(B)
D 50.73(a)(~)(viii)(B)
D 20.2203(a)(1J D 20.2203(al(4l D so.13(al(2l(iiil D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A) lg] 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x) 1 O. 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)(4l D 20.2203(a)(2)(iii)
D so.3s(c)(2)
D 50.73(a)(2)(v)(B)
D 73. 11 (a)(s)
D 20.2203(a)(2)(iv)
D 5o.4a(a)(3)(iil D 50.73(a)(2)(v)(C)
D 13.11(a)(1) 021 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 YEAR 2019 -
- 3. LEA NUMBER SEQUENTIAL
- NUMBER 001 D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES AND OPERATOR ACTIONS:
The dates, times, and major occurrences and operator actions for this event are as follows.
April 14, 2019 00:01:15-TSV-13 stroked close 00:01:41-TSV-13 opened 00:01 :44-AII Turbine Stop Valves closed 00:02:57 -Bypass Valves start to shut 00:03: 16-Reactor Scram on high reactor pressure 00:03: 19-Turbine trip 00:03:20-HPCI initiation signal due to Turbine Trip 00:03:52-HPCI reset
E. METHOD OF DISCOVERY
This event was-discovered by Reactor Operators when the reactor scram was received.
F. SAFETY SYSTEM RESPONSES:
All safety systems responded per design.
II. CAUSE OF EVENT
The cause of the scram is being investigated and will be provided in the supplement.
III. ANALYSIS OF THE EVENT
The automatic reactor scram and specified system activation is reportable under 1 O CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.73(a)(2)(iv)(A), as any event or condition that resulted in manual or automatic action of any of the specified systems listed in 10 CFR 50.73(a)(2)(iv)(B).
All other plant systems performed per design. Plant parameters, other than the RPV pressure, remained within normal values throughout the event. There was no loss of offsite power to the onsite emergency buses, the HPCI mode of feed and condensate system initiated as designed on the turbine trip signal.
REV NO.
00 YEAR 2019 *
- 3. LER NUMBER SEQUENTIAL NUMBER 001 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.
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 Restoring Arm Breakdown was returned to the proper position and tested satisfactorily.
B_ ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
To be provided in the supplement when the root cause evaluation is completed.
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
Turbine Stop Valve (TSV) -13, Restoring Arm Breakdown,
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:
COMPONENT Reactor Pressure Vessel Feedwater Level Control System High Pressure Coolant Injection System Reactor Protection System Turbine Stop Valve Main Turbine IEEE 803 FUNCTION IDENTIFIER RPV N/A N/A NIA PCV
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