05000293/LER-2016-007, Regarding Manual Reactor Scram Due to Feedwater Regulating Valve Malfunction
| ML16320A006 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/04/2016 |
| From: | Perkins E Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.16.067 LER 16-007-00 | |
| Download: ML16320A006 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(2) |
| 2932016007R00 - NRC Website | |
text
~~*Entergx November 4, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth. MA 02360
SUBJECT:
Licensee Event Report 2016-007-00, Manual Reactor Scram Due To Feedwater Regulating Valve Malfunction Pilgrim Nuclear Power Station Docket No. 50-293 Renewed License No. DPR-35 LEITER NUMBER: 2.16.067
Dear Sir or Madam:
The enclosed Licensee Event Report 2016-007-00, Manual Reactor Scram Due To Feedwater Regulating Valve Malfunction, is submitted in accordance with 10 CFR 50.73.
If you have any questions or require additional information, please contact me at (508) 830-8323. -
There are no regulatory commitments contained in this letter.
Sincerely,
~{?~~~~
Everett P. Perkins, Jr.
Manager, Regulatory Assurance EPP/sc
Attachment:
Licensee Event Report 2016-007-00, Manual Reactor Scram Due To Feedwater Regulating Valve Malfunction (4 Pages)
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station cc:
Mr. Daniel H. Dorman Regional Administrator, Region I U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 Ms. Booma Venkataraman, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8C2A Washington, DC 20555 NRC Senior Resident Inspector Pilgrim Nuclear Power Station Letter No. 2.16.067 Page 2 of 2
Attachment Letter Number 2.16.067 Licensee Event Report 2016-007-00 Manual Reactor Scram Due To Feedwater Regulating Valve Malfunction (4 Pages)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
... ~-;.,,_
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
\\~) ~1 LICENSEE EVENT REPORT (LER)
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections
.,,?...... *"'
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by (See Page 2 for required number of internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC digits/characters for each block) 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME [2. DOCKET NUMBER r* PAGE Pilgrim Nuclear Power Station 05000293 1 OF4
- 4. TITLE Manual Reactor Scram Due To Feedwater Regulating Valve Malfunction
- 5. EVENT DATE
- 6. LER 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 09 06 2016 2016
- - 007
- - 00 11 04 2016 N/A N/A
- 9. OEPRATING 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)(i)(C)
D 50.73(a)(2)(vii)
N D 20.2201(d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D so.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1)(ii)(A)
~ 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
D 20.2203(a)(2)(iii)
D so.36(c)(2)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4) 91 D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(C) 0 OTHER D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT I TELEPHONE NUMER (Include Area Code}
Mr. Everett P. Perkins, Jr. - Regulatory Assurance Manager 508-830-8323 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX B
SJ CON T351 No
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~NO SUBMISSION DATE
\\ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On September 6, 2016 at 0827 EDT, with the reactor at approximately 91 percent core thermal power, operators manually scrammed the reactor when the benchmark for a reactor water level of +42 inches increasing was reached. Following the scram, all rods fully inserted and the Average Power Range Monitors were downscale, indicating the reactor was shut down. The Main Steam Isolation Valves closed on a Primary Containment Isolation Signal Group 1 isolation and were subsequently reopened to maintain reactor pressure.
The reason for the increasing reactor water level was the malfunction of Feedwater Regulating Valve 'A' (FRV
'A'). The reactor operator at the control panel placed FRV 'A' in remote manual in an attempt to stabilize the feedwater flow oscillations. This had no effect on the performance of FRV 'A'. The operators experienced feedwater flow oscillations from FRV 'A' that resulted in reactor water level high and low level alarms on the control panel. Shortly after this, the benchmark reactor water level of +42 inches increasing was reached and operators manually scrammed the reactor.
There was no impact to public health and safety.
NRG FORM 366 (02-2014)
SAFETY CONSEQUENCES
- 3. PAGE 4 OF4 The actual consequences were a loss of ability to control FRV 'A' which resulted in increasing reactor water level and an operator-initiated manual scram. There were no other actual consequences to general safety of
~he public, nuclear safety, industrial safety, and radiological safety for this event.
The potential consequences to general safety of the public, nuclear safety, industrial safety, and radiological safety of this event if the operator-initiated manual scram did not occur were minimal. Had the manual scram not been initiated, in a short time there would have been an automatic turbine trip, feedwater pump trip, and Primary Containment Isolation Signal Group 1 isolation from the high reactor water level condition. These automatic actions would have immediately caused an automatic reactor Scram.
rThe conditional core damage probability of this event has been estimated to be 3.49E-06. This is the value associated with a Main Steam Isolation Valve (MSIV) isolation initiator. The risk from the actµal event was less than this calculated value because the Scram was manually initiated prior to the turbine trip and other automatic actions. The scram was considered uncomplicated as defined in Nuclear Energy Institute 99-02.
Therefore, based on the above there was no adverse impact on the public health or safety.
REPORT ABILITY ifhis report is submitted in accordance with 10 CFR 50. 73(a)(2)(iv)(A). Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section including 50.73(a)(2)(iv)(B)(2), general containment isolation signals c;iffecting containment isolation valves in more than one system or multiple MSIVs.
PREVIOUS EVENTS IA review of Pilgrim Nuclear Power Station Licensee Even't Reports for the past five years did not identify any
$imilar occurrences of manual scrams being initiated by feedwater oscillations.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EllS) CODES ifhe EllS codes for Components and Systems referenced in this report are as follows:
SYSTEMS CODES Feedwater System SJ
REFERENCES: