05000333/LER-2005-005, Re Automatic Reactor Scram on Low Reactor Vessel Water Level During Reactor Feed Pump Control Reset
| ML053220024 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/07/2005 |
| From: | Ted Sullivan Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-05-0172 LER 05-005-00 | |
| Download: ML053220024 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 3332005005R00 - NRC Website | |
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En tergy Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. Fitzpatrick NPP RO. Box 110 Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 T.A. Sullivan Site Vice President - JAF November 7, 2005 JAFP-05-0172 United States Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555
Subject:
Docket No. 50-333 LICENSEE EVENT REPORT: LER-05-005 (CR-JAF-2005-03818)
Automatic Reactor Scram on Low Reactor Vessel Water Level During Reactor Feed Pump Control Reset
Dear Sir:
This report is submitted in accordance with 10CFR50.73(a)(2)(iv)(A) based on the automatic actuation of systems listed in 10CFR50.73(a)(2)(iv)(B), including the Reactor Protection System, High Pressure Coolant Injection system, Reactor Core Isolation Cooling system and general containment isolation signals affecting more than one system (Group 2 isolation).
There are no commitments contained in this report.
Questions concerning this report may be addressed to Mr. Jim Costedio at (315) 349-6358.
TAS:DD:dd Enclosure cc:
USNRC, Region 1 USNRC, Project Directorate USNRC Resident Inspector INPO Records Center C
Abstract
On September 14, 2005, at approximately 0213, with the plant operating at 100% power, an automatic reactor scram on low reactor water level occurred following a momentary loss of the Uninterruptible Power Supply (UPS) system. The power loss resulted in a lockout of the Reactor Feed Pump (RFP) controls, as designed. Following reset of the RFP controls, a level transient occurred causing reactor water level to lower, resulting in an automatic reactor scram on low reactor pressure vessel (RPV) water level (Level l1l). A Group 2 Primary Containment Isolation System (PCIS) isolation occurred, resulting in multiple system isolations. The Reactor Core Isolation Cooling system (RCIC) and the High Pressure Coolant Injection system (HPCI) auto initiated on "low-low" RPV water level (Level II). All systems operated as designed during and after the reactor scram.
The level transient was caused by a RFP controller low output signal which is a design characteristic of the controller following a momentary loss of power. The operator did not verify the output signal prior to resetting the RFP control, which would have identified the low signal. Contributing to this human error was an inadequate abnormal operating procedure. As part of the corrective actions, the applicable procedure has been revised and the requirement to verify the controller output signal prior to resetting has been reinforced.
There were no safety system functional failures or other safety consequences associated with this event.
NRC FORM 366 (6-2004)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION 2
OF 4
NUMBER NUMBER James A. FitzPatrick Nuclear Power Plant 05000333 05 005 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (11 7)
EIIS Codes in I I PLANT STATUS The James A. FitzPatrick Nuclear Power Plant (JAF) was in Mode 1 (Run) at 100% power at the time of the automatic reactor scram. Just prior to the event, both Reactor Feed Pumps (RFPs) [SK] were operating at approximately 4000 revolutions per minute (RPM) and the plant was preparing for planned maintenance on the Uninterruptible Power Supply (UPS) motor generator set.
BACKGROUND Feedwater [SJ] is delivered to the reactor pressure vessel (RPV) by two turbine driven RFPs. The two RFPs are single-stage, horizontal, centrifugal units using a steam driven turbine for motive power. The pumps operate in series with the condensate and condensate booster pumps to provide the required design flow and pressure at the reactor inlet nozzles.
The Feedwater Control System (FCS) [JB] controls the RFPs to automatically regulate feedwater flow into the RPV. The FCS is also capable of being manually regulated. Upon a loss of feedwater control signal, such as during a momentary loss of power, the control system voltage is removed from the RFP speed control gear, which locks the RFP speed at the speed level demanded just prior to the loss of control signal. The RFP speed can then be manually controlled or the control lockout can be reset to return the RFP speed control to automatic.
EVENT DESCRIPTION
On September 14, 2005, at approximately 0213, with the plant operating at 100% power, an automatic reactor scram on low RPV water level (Level ll) occurred as a result of both RFPs decreasing in speed.
Just prior to this event, during preparations for planned maintenance on the UPS motor generator set, a momentary loss of the UPS system occurred while transferring UPS loads to the alternate power supply. The power loss resulted in a loss of feedwater control signal causing a lockout of the RFP controls, as designed. The appropriate Abnormal Operating Procedure (AOP) -21, "Loss of UPS", was entered. The AOP directs an operator to reset the "A" and "B" RFP loss of signal lockouts to restore automatic RPV water level control. A Senior Nuclear Operator (SNO) performed this action without verifying the associated output signals of the RFP controls. Due to the feedwater control output being low at the time of reset, both RFP's decreased in speed from approximately 4000 RPM to approximately 2500 RPM. This decrease in speed caused the RPV water level to decrease to <177" above Top of Active Fuel (TAF), the 'low" level scram setpoint (Level l1l),
causing the Reactor Protection System (RPS) to initiate an automatic reactor scram.
AOP-1, "Reactor Scram", and the associated Emergency Operating Procedure (EOP) -2, "RPV Control", were promptly entered and executed.
A Group 2 Primary Containment Isolation System (PCIS) [JM] isolation occurred due to RPV water level falling below Level III, resulting in multiple system isolations. The Reactor Core Isolation Cooling system (RCIC) [BN] and the High Pressure Coolant Injection system (HPCI).[BJ] auto initiated on "low-low" RPV water level (Level II, <126.5" above TAF), as designed.
The HPCI system did not inject into the RPV due to the prompt recovery of RPV water level, which caused the HPCI injection valve to remain closed. The RCIC system started, aligned itself for injection, and injected into the RPV. Both HPCI and RCIC initiated and tripped appropriately based upon changing reactor water level conditions. The response of both systems was as expected.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL I REVISION 3
OF 4
NUMBER NUMBER James A. FitzPatrick Nuclear Power Plant 05000333 05 005 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (I 7)All systems responded as designed during and after the reactor scram. All control rods fully inserted. The post transient evaluation determined that appropriate operator response was demonstrated during scram recovery activities. There were no nuclear, radiological or personnel safety issues associated with this event.
BASIS FOR REPORT This event is reportable under 10CFR50.73(a)(2)(iv)(A) based on the automatic actuation of RPS, PCIS, HPCI and RCIC.
CAUSE OF EVENT
The cause of the reactor scram was low RPV water level resulting from an unanticipated reduction in RFP speed. This reduction in RFP speed was the result of a RFP control low output signal at the time the control lockout was reset. A proper verification of the control output signal would have identified this control low output prior to resetting the control lockout. This verification is a station expectation and is reinforced during operator training, but was not performed by the responsible SNO. [Cause Code A]
Insufficient guidance in the associated AOP contributed to this event. The applicable AOP (AOP-21) was reviewed during the pre-job brief. However, the AOP did not require the operator to observe the control outputs prior to returning the system to automatic operation and required the feedwater system lockouts to be reset promptly. This resulted in the SNO focusing on the action of promptly resetting the control lockouts instead of first verifying RFP control output.
EVENT ANALYSIS
The plant responded as designed following the automatic reactor scram. There were no challenges to the reactor coolant pressure boundary or the fuel cladding integrity. This event and the transient response is bounded by previous analyses contained in the Final Safety Analysis Report (FSAR), including various generation load reject transients, turbine trip transients, and reactor isolation transients. Therefore, the safety significance of this event was low and there were no nuclear, radiological or personnel safety issues associated with this event.
CORRECTIVE ACTIONS
Corrective Actions Completed Prior to this Report:
- 1.
Revised the applicable AOP (AOP-21) to eliminate the requirement for promptly resetting the RFP control lockouts subsequent to a momentary loss of UPS.
Corrective Actions not yet Completed:
- 1.
Conduct operator training on the event and on the RFP controller response.
(Due 12/15/2005)
- 2.
Perform extent of condition assessment for operator errors involving self checking during equipment manipulation.
(Due 11/15/2005)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION4 OF 4
NUMBER NUMBER James A. FitzPatrick Nuclear Power Plant 05000333 05 005 00 TEXT (If more space is required, use additional copies of NRC Form 366AJ 117)
SAFETY SYSTEM FUNCTIONAL FAILURE REVIEW A review of this event determined that no safety system functional failure occurred.
SIMILAR EVENTS
No other similar issues were identified in previous plant LERs.
FAILED COMPONENT IDENTIFICATION There were no component failures that directly caused this event.
REFERENCES
- 1. Root Cause Analysis Report, JAF Condition Report CR-JAF-2005-03818, Automatic Reactor Scram on Low Reactor Vessel Water Level, dated October 12, 2005.