05000220/LER-2012-007, Regarding High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level
| ML13158A301 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/03/2013 |
| From: | Philippon M Constellation Energy Nuclear Group, EDF Group, Nine Mile Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 12-007-00 | |
| Download: ML13158A301 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(iii) 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 2202012007R00 - NRC Website | |
text
Michel A. Philippon P.O. Box 63 Plant General Manager Lycoming, New York 13093 315.349.5205 315.349.1321 Fax CENG.
a joint venture of Consteflation ey:eDF NINE MILE POINT NUCLEAR STATION June 3, 2013 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:
Document Control Desk
SUBJECT:
Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 Licensee Event Report 2012-007, High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level In accordance with 10 CFR 50.73(a)(2)(iv)(A), please find attached Licensee Event Report (LER) 2012-007, High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level. The event notification per 10 CFR 50.72(b)(3)(iv)(A) for the High Pressure Coolant Injection (HPCI) system actuation was initially completed on November 6, 2012 (Event Number 48481). This notification was subsequently retracted on December 17, 2012, on the basis that the actuation was invalid. Following further review, on April 24, 2013, Nine Mile Point Nuclear Station, LLC (NMPNS) confirmed that the event did constitute a valid actuation of the HPCI system and was reportable per 10 CFR 50.72(b)(3)(iv)(A). NMPNS recognizes that, based on the initial date of occurrence of the event, submittal of this LER is not timely.
There are no regulatory commitments in this submittal.
Should you have questions regarding the information in this submittal, please contact John J. Dosa, Director Licensing, at (315) 349-5219.
Very truly yours, MAP/DEV
Document Control Desk June 3, 2013 Page 2
Attachment:
Licensee Event Report 2012-007, High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level cc:
Regional Administrator, NRC Project Manager, NRC Resident Inspector, NRC
ATTACHMENT LICENSEE EVENT REPORT 2012-007 HIGH PRESSURE COOLANT INJECTION SYSTEM LOGIC ACTUATION FOLLOWING AN AUTOMATIC TURBINE TRIP SIGNAL DUE TO HIGH REACTOR WATER LEVEL Nine Mile Point Nuclear Station, LLC June 3, 2013
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-20 10)
, the NRC 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 6
- 4. TITLE High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED I
I FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR None NA NUMBER NO.NoeA 11 OPRAIN M
0DE 201 00 00 060-21 FACILITY NAME DOCKET NUMBER 12 2
003l None NA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[O 20.2201(b)
[I 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
N [1 20.2201(d)
[I 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[E 20.2203(a)(1)
C1 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
- 10. POWER LEVEL
'[ 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
[I 20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 000 El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
[E 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[I 50.73(a)(2)(i)(B)
[I 50.73(a)(2)(v)(D)
Specify in Abstract below or in There was no impact on Nine Mile Point Unit 2 from this event.
The event notification per 10 CFR 50.72(b)(3)(iv)(A) for the HPCI system actuation was initially completed on November 6, 2012 at 0356 hours0.00412 days <br />0.0989 hours <br />5.886243e-4 weeks <br />1.35458e-4 months <br /> (Event Number 48481). This notification was subsequently retracted on December 17, 2012, on the basis that the actuation was invalid.
Following further review, on April 24, 2013, Nine Mile Point Nuclear Station, LLC (NMPNS) confirmed that the event did constitute a valid actuation of the HPCI system and was reportable per 10 CFR 50.72(b)(3)(iv)(A). NMPNS recognizes that, based on the initial date of occurrence of the event, submittal of this LER is not timely.
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
Prior to the event, there were no inoperable structures, systems, or components that contributed to the event.
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
On November 6, 2012, the 12 Feedwater flow control valve (FCV-29-137) opened unexpectedly, causing reactor water level to rise. An automatic turbine trip signal on high reactor water level occurred at 0006 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which then resulted in actuation of the HPCI channels 11 and 12 initiation logic, by design. The operators closed the 12 Feedwater pump discharge blocking valve to terminate the rise in reactor water level.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
None.
F. METHOD OF DISCOVERY
This event was discovered by operator observation of control room indication of rising reactor water level following application of the tagout for the performance of maintenance on the reactor feedwater level control circuitry.
G. MAJOR OPERATOR ACTION:
Upon observing the rising reactor water level indication in the control room, the operators tripped the 11 CRD pump and closed the 12 Feedwater pump discharge blocking valve (VLV-29-09) to control reactor water level. The operators stabilized reactor water level by using reject flow from the reactor water cleanup system and by re-starting the 11 CRD pump.
H. SAFETY SYSTEM RESPONSES:
The HPCI system initiation logic actuated in response to the automatic turbine trip signal, as designed. No other safety system responses occurred or were required as a result of this event.
NRC FORM 366 (10-2010)
II. CAUSE OF EVENT
The cause of this event is failure to use adequate human performance tools (specifically, a questioning attitude) when performing periodic testing of the feedwater flow control valves in accordance with procedure N1-IPM-029-010, "Calibration of Feedwater FCV-29-134, FCV-29-137, and FCV-29-141." The procedure tests the functioning of the lockup valves by applying air pressure to the top and bottom cylinders of the flow control valve actuator and then determining the amount of air pressure decrease over a 5-minute time period. A pressure decrease of 20 percent or more in either cylinder indicates that the lockup valve should be re-built. The last time that this test was performed in March 2011, the test was performed improperly in that the lockup valve pressure drop test was conducted without the actuating cylinder being pressurized. The test personnel did not question the test results when zero test pressure was measured. Thus, since performance of procedure Nl-IPM-029-010 did not detect the degraded lockup valve o-rings, no corrective actions were taken. In addition, there is no preventive maintenance activity to re-build the lockup valves at any given frequency; thus, the lockup valve o-ring degradation was not identified and corrected prior to the event that occurred on November 6, 2012.
This event was entered into the NMPNS corrective action program as condition report number CR-2012-010141.
Ill. ANALYSIS OF THE EVENT:
This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B). The NMP1 HPCI system is a feedwater coolant injection system, which is one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B).
There were no actual safety consequences associated with this event. The unexpected opening of the 12 Feedwater flow control valve (FCV-29-137) resulted in a rising reactor water level that caused a turbine trip signal, which then resulted in actuation of the HPCI channels 11 and 12 initiation logic, by design. HPCI is a mode of operation that utilizes selected equipment of the condensate and feedwater system to perform its function. The HPCI system is not an emergency core cooling system and is not considered in any loss of coolant accident analyses. It is available to provide core cooling in the event of a small reactor coolant line break which exceeds the capability of the CRD pumps.
At the time of the event on November 6, 2012, with the reactor in cold shutdown, the main turbine was not in service and the HPCI system was not required by the NMP1 Technical Specifications to be operable. Though the HPCI initiation logic was actuated, no HPCI components actually started or actuated, and HPCI system injection into the reactor vessel neither occurred nor was required, since adequate core cooling was already being provided. Plant parameters other than reactor water level remained within normal values throughout the event.
Based on the above, it is concluded that the actual 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.
NRC FORM 366 (10-2010)
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
Immediate actions were taken by the operators to terminate the rise in reactor water level and to restore reactor water level to the operating band. With the plant already in the cold shutdown condition, no further actions were required.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
- 1. The rubber goods in the lockup valves for the 11 and 12 Feedwater flow control valves (FCV-29-141 and FCV-29-137) were replaced.
- 2. Instrumentation and Controls maintenance personnel were briefed on this event, including the importance of the continual use of human performance tools when performing maintenance tasks.
- 3. A change to procedure Nl-IPM-029-010 was processed to clearly define the minimum starting air pressure required prior to commencing the lockup valve test.
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
The lockup valves (BV-29-231 and BV-29-232) for Feedwater flow control valve FCV-29-137.
B. PREVIOUS LERs ON SIMILAR EVENTS:
There have been several previous LERs for events involving a turbine trip and subsequent actuation of the HPCI system initiation logic (LERs 2006-002, 2009-002, and 2012-005). The causes and actions described in these previous LERs were different than the current event and would not have prevented this event.
NRC FORM 366 (10-2010)
C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
COMPONENT Feedwater System Feedwater Flow Control Valve Feedwater Isolation Valve High Pressure Coolant Injection System Main Turbine/Supervisory Control Reactor Vessel Control Rod Drive Pump IEEE 803 COMPONENT IDENTIFIER FCV ISV TRB RPV P
IEEE 805 SYSTEM IDENTIFICATION SJ SJ SJ SJ JJ AD AA D. SPECIAL COMMENTS:
None NRC FORM 366 (10-2010)